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Sample records for c reactor

  1. Prometheus Reactor I&C Software Development Methodology, for Action

    Energy Technology Data Exchange (ETDEWEB)

    T. Hamilton

    2005-07-30

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I&C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I&C Software Development Process Manual and Reactor Module Software Development Plan to NR for information.

  2. Prometheus Reactor I and C Software Development Methodology, for Action

    International Nuclear Information System (INIS)

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I and C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I and C Software Development Process Manual and Reactor Module Software Development Plan to NR for information

  3. Upgrading I and C for the Es Salam research reactor

    International Nuclear Information System (INIS)

    The Es Salam is a multi-purpose research reactor intended for the production of the radio elements, material tests, education and training. It serves also as a source for neutron beams used by chemists, biologists, metallurgists and physicists for fundamental research and applications. The main task of the Es Salam Reactor Operation Division is to ensure a safe and reliable operation of the reactor. With this objective in hand, the Operation Division is responsible for updating the safety measures and conditions in the installation for the reactor equipments and systems. Due to the increased demand for experiences and to the ageing effects, modification and modernization of some safety items become necessary in comparison with state of the art installations. Furthermore, the technological advances and the development and introduction of new instruments, components and systems increased the need for modification. In this paper, we present the new I and C system which will be used for the Es Salam research reactor

  4. Justification for extending scram recovery time at C Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Owsley, G.F.

    1956-07-13

    Slug ruptures at C Reactor are costing a considerable amount of production and are limiting reactor power levels. A large portion of this production is lost because the ``quickie`` procedure cannot be used with unstruck ruptures. Currently there are two reasons for this condition: Insufficient reactivity is available to recover after a rupture is removed with the ``quickie`` equipment, and if the ``quickie`` method were attempted, there would be many failures due to stuck rear caps and improperly functioning equipment. With the current emphasis being placed on improving the ``quickie`` equipment and procedures, it is reasonable to assume that in the near future nearly all unstruck ruptures can be removed with this ``quickie`` procedure. However, the C Reactor does not possess sufficient excess reactivity to recover from a ``quickie`` rupture discharge following a shutdown from equilibrium conditions. This study outlines a procedure for extending the scram recovery time at C Reactor and an economic justification for buying the added reactivity necessary to carry out the procedure.

  5. Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O2 cannot be ignored, especially for the FHR, in which environment the product, SiO2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

  6. Three-dimensional analysis of SiC/SiC composite structures for fusion reactors

    International Nuclear Information System (INIS)

    The structural analysis of a SiC/SiC composite first wall in a tokamak fusion reactor is explored using the finite element method. The analysis is done in three dimensions under thermal and pressure loads. The effect of the three-dimensional analysis on the results is compared with that for the two-dimensional analysis. The comparison indicates that for laminated SiC/SiC composite structures, the two-dimensional analysis, which ignores the effect of the stacking sequence of the laminate and the effect of the interlaminar shear stress, is insufficient for reactor design. 9 refs., 14 refs., 2 tabs

  7. Characterization of 14C in Swedish light water reactors.

    Science.gov (United States)

    Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

    2008-08-01

    This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units. PMID:18617793

  8. Basic safety principles of KLT-40C reactor plants

    International Nuclear Information System (INIS)

    The KLT-40 NSSS has been developed for a floating power block of a nuclear heat and power station on the basis of ice-breaker-type NSSS (Nuclear Steam Supply System) with application of shipbuilding technologies. Basic reactor plant components are pressurised water reactor, once-through coil-type steam generator, primary coolant pump, emergency protection rod drive mechanisms of compensate group-electromechanical type. Basic RP components are incorporated in a compact steam generating block which is arranged within metal-water shielding tank's caissons. Domestic regulatory documents on safety were used for the NSSS design. IAEA recommendations were also taken into account. Implementation of basic safety principles adopted presently for nuclear power allowed application of the KLT-40C plant for a floating power unit of a nuclear co-generation station. (author)

  9. 105-C Reactor interim safe storage project technology integration plan

    International Nuclear Information System (INIS)

    The 105-C Reactor Interim Safe Storage Project Technology Integration Plan involves the decontamination, dismantlement, and interim safe storage of a surplus production reactor. A major goal is to identify and demonstrate new and innovative D and D technologies that will reduce costs, shorten schedules, enhance safety, and have the potential for general use across the RL complex. Innovative technologies are to be demonstrated in the following areas: Characterization; Decontamination; Waste Disposition; Dismantlement, Segmentation, and Demolition; Facility Stabilization; and Health and Safety. The evaluation and ranking of innovative technologies has been completed. Demonstrations will be selected from the ranked technologies according to priority. The contractor team members will review and evaluate the demonstration performances and make final recommendations to DOE

  10. Upgrade of VR-1 training reactor I and C

    International Nuclear Information System (INIS)

    The contribution describes the upgrade of the VR-1 training reactor I and C (Instrumentation and Control). The reactor was put into operation in the 1990, and its I and C seems to be obsolete now. The new I and C utilises the same digital technology as the old one. The upgrade has been done gradually during holidays in order not to disturb the reactor utilisation during teaching and training. The first stage consisted in the human-machine interface and the control room upgrade in 2001. A new operator's desk, displays, indicators and buttons were installed. Completely new software and communication interface to the present I and C were developed. During the second stage in 2002, new control rod drivers and safety circuits were installed. The rod motors were replaced and necessary mechanical changes on the control rod mechanism, induced by the utilisation of the new motor, were done. The new safety circuits utilise high quality relays with forced contacts to guarantee high reliability of their operation. The third stage, the control system upgrade is being carried out now. The new control system is based on an industrial PC mounted in a 19 inch crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. A large amount of work has been devoted to the software requirements to specify all dependencies, modes and permitted actions, safety measures, etc. The Department took an active part in the setting of software requirements and later in verification and validation of the software and the whole control system. Finally, a new protection system consisting of power measuring and power protection channels will be installed in 2004 or 2005. (author)

  11. STAR-C Space Thermionic Advanced Reactor - Compact

    International Nuclear Information System (INIS)

    STAR-C is a small, compact nuclear space power system that can be configured to provide 5 to 25 kW of electrical power for long duration space applications. A 10 kWe baseline design concept has been defined that has a 550 kg mass and is one meter long and 2/3 meter in diameter. The reactor configuration is based on the Soviet ''ROMASHKA'' reactor that was built and operated in the early 1960's. The thermionic power conversion as based on the U.S. Solar Energy Thermionic (SET) program. The reactor fuel is derived from the NERVA nuclear rocket program. A number of military spacecraft are under development for deployment in the 1990's. Power requirements for many of these systems range from 5 to 25 kWe. Typically, these platforms require low levels of power for continuous housekeeping functions and higher levels for alert and battle engagement conditions. It is highly desirable that the selected space power system have a substantial growth capability that can be utilized as the spacecraft concept matures. There is also an increased emphasis on requirements for system survivability to hostile weapons threats and the capability for spacecraft maneuverability for evasive action. This requires that the space power system must be compact and that it must be closely integrated into the spacecraft structure to avoid inertial and dynamic effects associated with the use of extendable structures and booms

  12. Design of first reactor protection system prototype for C A R E M reactor

    International Nuclear Information System (INIS)

    In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility

  13. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 2000C. The design description and results of the prototype capsule performance are presented

  14. Gas permeability of SiC/SiC composites as fusion reactor material

    International Nuclear Information System (INIS)

    Permeability of helium gas in SiC/SiC composites material, which is one of the most important properties in application of SiC/SiC composite for first wall and blanket of fusion reactors, was studied by using a vacuum apparatus. Three tubular and two flat plate SiC/SiC composites were prepared by different preparation processes. The measurement of permeability coefficient of helium gas was carried out with pressure ranging from 102 to 105 Pa at room temperature. The permeability coefficient of the SiC/SiC composite largely depended on the preparation method. In three tubular materials, the SiC/SiC composite made by both polymer impregnation and pyrolysis (PIP) and melt infiltration (MI) methods showed the lowest permeability, 9.1x10-7 m2/s, which was approximately two orders of magnitude smaller than one of the material made only by PIP method. The permeability of the flat plate SiC/SiC composites made by both liquid phase sintering (LPS) and hot pressing (HP) was approximately 1.5x10-9-4.0x10-11 m2/s. The difference of permeability was related to the microscopic structure, i.e. pores and cracks

  15. A new SiC/C bulk FGM for fusion reactor

    International Nuclear Information System (INIS)

    Graphite is widely used in present Tokamak facilities and a C/C composite has been selected as one of the candidate materials for the ITER. But C-based material has an excessive chemical sputtering yield at 600-1000 K and exhibits irradiation enhanced sublimation at >1200 K under plasma erosion condition, causing serious C-contamination of plasma. Low Z material SiC has several advantages for use in fusion reactor, such as excellent high temperature properties, corrosion resistance, low density, and especially its low activation irradiation. To reduce C contamination during plasma exposure, previously SiC coatings were chemically deposited on the surface of C-substrate, however, the thermal stresses arise on the interface between the coating layers and the substrate under high temperature. Heating/cooling cycle leading to cracks in SiC/C interface, small thickness of coating and long processing time are limiting factors for FGM made with CVD process. In this paper, a new SiC/C bulk FGM has been successfully fabricated with P/M hot pressing process. The chemical sputtering yield, gas desorption performance, thermal shock resistance and physical sputtering performance in Tokamak are outlined in this paper. (author)

  16. Reliability Analysis of I and C Architecture of Research Reactors Using Bayesian Networks

    International Nuclear Information System (INIS)

    The objective of this research project is to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and BN models were constructed. Bayesian network analysis was performed to find the reliability features. This is a continuation of study towards the standardization of I and C architecture for low and medium power research reactors. This research is the continuation of study to analyze the reliability of single channel of Reactor Protection System (RPS) using Bayesian networks. The focus of research was on the development of architecture for low power research reactors. What level of reliability is sufficient for protection, safety and control systems in case of low power research reactors? There should be a level which should satisfy all the regulatory requirements as well as operational demands with optimized cost of construction. Scholars, researchers and material investigators from educational and research institutes are demanding for construction of more research reactors. In order to meet this demand and construct more units, it is necessary to do more research in various areas. The research is also needed to make a standardization of research reactor I and C architectures on the same lines of commercial power plants. The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250-2.0 MW power rating or 2.5-10 Χ 1011 n/cm2.s. flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 1013 n/cm2.s. are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor

  17. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R.

    1976-07-01

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  18. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte I

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-07-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  19. Design requirements for SiC/SiC composites structural material in fusion power reactor blankets

    International Nuclear Information System (INIS)

    This paper recalls the main features of the TAURO blanket, a self-cooled Pb-17Li concept using SiC/SiC composites as structural material, developed for FPR. The objective of this design activity is to compare the characteristics of present-day industrial SiC-SiC composites with those required for a fusion power reactor blanket (FPR) and to evaluate the main needs of further R and D. The performed analyses indicated that the TAURO blanket would need the availability of SiC/SiC composites approximately 10 mm thick with a thermal conductivity through the thickness of approximately 15 Wm-1K-1 at 1000 C and a low electrical conductivity. A preliminary MHD analysis has indicated that the electrical conductivity should not be greater than 500 Ω-1m-1. Irradiation effects should be included in these figures. Under these conditions, the calculated pressure drop due to the high Pb-17Li velocity (approximately 1 m s-1) is much lower then 0.1 MPa. The characteristics and data base of the recently developed 3D-SiC/SiC composite, Cerasep trademark N3-1, are reported and discussed in relation to the identified blanket design requirements. The progress on joining techniques is briefly reported. For the time being, the best results have been obtained using Si-based brazing systems initially developed for SiC ceramics and whose major issue is the higher porosity of the SiC/SiC composites. (orig.)

  20. Gas permeability of SiC/SiC composite as blanket material of fusion reactor

    International Nuclear Information System (INIS)

    Gas permeability of SiC/SiC composite materials, which is one of the most important properties in application of SiC/SiC composite for first wall and blanket of fusion reactors, was measured by using a vacuum apparatus. The cylindrical SiC/SiC composite specimens were prepared by three different processes. The measurement on permeability for three materials was carried out with helium gas pressure ranging from 102 to 105 Pa at room temperature. The pressure in bottom chamber down stream of specimens increased with the helium gas pressure within the applied pressure range. The helium gas flow through the material is regarded as molecular flow. The material made by PIP method showed the highest permeability. The lowest permeability was observed in the one made by PIP followed by RS method. The material, SA-TyrannoHexTM made by hot pressing was in the second position. The difference of the permeability can be related with the macroscopic structure represented by pores and cracks. (author)

  1. C and I power supply for research reactors (Paper No. CP 33)

    International Nuclear Information System (INIS)

    The safety and availability of a nuclear reactor depends greatly on the quality and reliability of control and instrumentation (C and I) power supply. It may not be practical to incorporate fail-safe philosophy in all safety systems of a reactor due to fact that one has to pay huge penalty for spurious actuation of certain systems. A reliable power supply system can solve this problem. Availability of the status of reactor and various systems is essential for the safe operation of the plant. In small size reactors the C and I is powered from normal power supply. It is not practical to provide battery power backup for the whole C and I due to charging and maintenance problems associated with large batteries. This is more so since the reactor is not operated on a continuous basis. But it becomes mandatory to provide some amount of battery power atleast to indicate certain vital information like reactor status etc. in case of failure of normal power supply. This paper describes various aspects of C and I power supply for the research reactors DHRUVA, PURNIMA and KAMINI and highlights their evolution in order to ensure safe and reliable operation of the reactor. (author). 6 refs

  2. Application of non-destructive testing and in-service inspections to research reactors and preparation of ISI programme and manual for WWR-C research reactors

    International Nuclear Information System (INIS)

    The present report gives a review on the results of application of non-destructive testing and in-service inspections to WWR-C reactors in different countries. The major problems related to reactor safety and the procedure of inspection techniques are investigated to collect the experience gained from this type of reactors. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of their rehabilitation programmes. 9 figs., 4 tabs

  3. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellarators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transport behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  4. Studies of a modular advanced stellarator reactor ASRA6C

    International Nuclear Information System (INIS)

    This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellerators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transort behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

  5. Experience with modernization and refurbishment of the Vienna TRIGA Mark II reactor I and C system

    International Nuclear Information System (INIS)

    The refurbishment of the instrumentation and control (I and C) system of a research reactor is a major task which needs careful planning and taking many aspects into account. At any early planning stage, the future of the facility has to be demonstrated to the national authorities by providing a detailed business plan and the cost of I and C replacement will be compared by financial authorities against the cost of decommissioning the facility. The TRIGA reactor Vienna was modernized in 1992 with a new digital instrumentation and control (I and C) system. The replacement procedure and the reactor-specific modifications to the standard reactor instrumentation offered by the supplier, the operation experience during the past 15 years and a compilation of benefits and other issues to be considered in these procedures (changing from analog to digital I and C system) are summarized in this report. (nevyjel)

  6. Project management plan for Reactor 105-C Interim Safe Storage project

    International Nuclear Information System (INIS)

    Reactor 105-C (located on the Hanford Site in Richland, Washington) will be placed into an interim safe storage condition such that (1) interim inspection can be limited to a 5-year frequency; (2) containment ensures that releases to the environmental are not credible under design basis conditions; and (3) final safe storage configuration shall not preclude or significantly increase the cost for any decommissioning alternatives for the reactor assembly.This project management plan establishes plans, organizational responsibilities, control systems, and procedures for managing the execution of Reactor 105-C interim safe storage activities to meet programmatic requirements within authorized funding and approved schedules

  7. Conceptual Design of I and C Architecture for a New Research Reactor

    International Nuclear Information System (INIS)

    NRR) by 2016 was launched by KAERI in 2012. The purposes of the project are to meet domestic industrial needs of a research reactor and to secure an internationally competitive NRR. For instrumentation and control (I and C) systems of the NRR, it shall also be designed to secure the competition of the NRR. The I and C should account for the competition in terms of three aspects: safety, performance and cost. A I and C conceptual design activity should be carried out at the early stage of the project to figure out the I and C on the high level. Constructing I and C architecture is to accomplish the high level design. During the I and C architecture construction, the three aspects should be taken into account. This paper conceptually constructs the I and C architecture for the NRR by comparing to the cases of a Jordan training and research reactor (JRTR) project and a RA-10 multipurpose research reactor. The JRTR is an on-going project launched by KAERI and DAEWOO consortium in 2011. The I and C of the JRTR was digitalized based on the I and C functions of the HANARO research reactor, as shown in Fig. 1. The RA-10 was launched by INVAP of Argentina in 2011. The digital I and C developed by INVAP was built in an OPAL of Australia, as shown in Fig. 2

  8. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  9. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I

    International Nuclear Information System (INIS)

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  10. Neutronics and thermohydraulics of the reactor C.E.N.E. Part II

    International Nuclear Information System (INIS)

    In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

  11. Phospholipase C-catalyzed sphingomyelin hydrolysis in a membrane reactor for ceramide production

    DEFF Research Database (Denmark)

    Zhang, Long; Liang, Shanshan; Hellgren, Lars; Jonsson, Gunnar Eigil; Xu, Xuebing

    2008-01-01

    A membrane reactor for the production of ceramide through sphingomyelin hydrolysis with phospholipase C from Clostridium perfringens was studied for the first time. Ceramide has raised a large interest as an active component in both pharmaceutical and cosmetic industry. The enzymatic hydrolysis o...... study demonstrated the improved enzyme reusability, the fast immobilization process, the straightforward up-scaling and the combination of the hydrolysis with the product separation in the membrane reactor developed....

  12. ATWA Frequency for the Analog I and C System of the OPR-1000 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Seungcheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) that results in a rapid pressure rise of the primary side by no reactor trip. The magnitude and timing of the reactor coolant system (RCS) pressure rise depends on the moderator temperature coefficient (MTC), the pressure relief capacity and the energy removal capacity of the secondary side in the pressurized water reactor (PWR). It is dealt with an important safety issue in the point that the primary pressure over ASME stress C level (3,200psig) can lead to core damage consequently. ATWS risk is simply defined as the multiplication of the ATWS frequency and unfavorable exposure time (UET). This paper focuses the estimation of an ATWS frequency for the OPR-1000 reactor with an analog reactor protection system (RPS). It is an important issue in risk-informed technical specification (RITS) of RPS. The plant-specific ATWS frequency model for the OPR-1000 reactor was developed using more realistic information and the state-of-art technology. The results of the work can be directly used to improve risk-informed surveillance test interval (RI-STI) of the KSNP safety-related I and C systems such as RPS.

  13. Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

    International Nuclear Information System (INIS)

    If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

  14. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  15. Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

    International Nuclear Information System (INIS)

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department

  16. Reactor pressure vessel steels ASME SA533B and SA508 C1.2

    International Nuclear Information System (INIS)

    The report presents the results of the microstructural studies of steels SA533B and SA508 C1.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of structural integrity of the reactor pressure vessels. The as-quenched and variably tempered microstructures were studied with optical, scanning and transmission electron microscopes. (author)

  17. Evaluation of 14C Behavior Characteristic in Reactor Coolant from Korean PWR NPP's

    International Nuclear Information System (INIS)

    This study has been focused on determining the chemical composition of 14C - in terms of both organic and inorganic 14C contents - in reactor coolant from 3 different PWR's reactor type. The purpose was to evaluate the characteristic of 14C that can serve as a basis for reliable estimation of the environmental release at domestic PWR sites. 14C is the most important nuclide in the inventory, since it contributes one of the main dose contributors in future release scenarios. The reason for this is its high mobility in the environment, biological availability and long half-life (5730 yr). More recent studies - where a more detailed investigation of organic 14C species believed to be formed in the coolant under reducing conditions have been made - show that the organic compounds not only are limited to hydrocarbons and CO. Possible organic compounds formed including formaldehyde, formic acid and acetic acid, etc. Under oxidizing conditions shows the oxidized carbon forms, possibly mainly carbon dioxide and bicarbonate forms. Measurements of organic and inorganic 14C in various water systems were also performed. The 14C inventory in the reactor water was found to be 3.1 GBq/kg in PWR of which less than 10% was in inorganic form. Generally, the activity in the water was divided equally between the gas- and water- phase. Even though organic compound shows that dominant species during the reactor operation, But during the releasing of 14C from the plant stack, chemical forms of 14C shows the different composition due to the operation conditions such as temperature, pH, volume control tank venting and shut down chemistry.

  18. Change of I-V characteristics of SiC diodes upon reactor irradiation

    International Nuclear Information System (INIS)

    In search for semiconductors, which can be used in high-flux reactors in order to measure flux distributions, we irradiated SiC p-n junctions in the Belgium BR-1 reactor. Two types of SiC-diodes of different origin have been irradiated. These junctions are grown in the Lely-furnace. The change in forward and reverse characteristics have been measured during and after irradiation up to temperatures of 150oC, while measurements up to a temperature of 500oC are in progress. It has been found that one type resists BR-1 neutrons up to an integrated flux of 1015 n/cm2, while the other resists irradiation up to a flux of 1017 n/cm2. The changes in characteristics are given as well as the result of some annealing experiments. (author)

  19. Production and release of 14C from a swimming pool reactor

    International Nuclear Information System (INIS)

    The annual production rate of 14C in the Apsara swimming pool reactor works out to be about 2.94 mCi. The concentration distribution of 14C in different compartments viz. pool water, reactor hall air and ion-exchange resin ranged from 200 to 440 pCi/l, 0.09 to 0.38 pCi/l, an average concentration of 8.16 pCi/g respectively. The mean residence time of 14C in pool water is evaluated to be about 7 days taking into account various sinks. The study revealed atmospheric exchange at the air-water interface as the dominant process responsible for the loss of 14C from the pool water. (author). 7 refs., 2 figs., 4 tabs

  20. A very simple facility for reactor irradiations from 70 - 4000C

    International Nuclear Information System (INIS)

    An irradiation loop was installed at the Research Reactor of 'Demokritos'. This facility allows the irradiation with fast neutrons from 70 to 4000C. A description is given of the loop. As an example of its use, a figure is shown of the negative resistivity curve of Cu + 50 at%Ni as a function of neutron dose at an irradiation temperature of 730C. (U.K.)

  1. Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

    OpenAIRE

    Diez de la Obra, Carlos Javier; Stankovskiy, Alexey; Malambu, E.; Zerovnik, Gasper; Schillebeeckx, Peter; Van Den Eynde, Gert; Heyse, Jan; Cabellos de Francisco, Oscar Luis

    2013-01-01

    A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled re...

  2. Irradiations in swimming-pool type reactors from room temperature up to 2000 deg C

    International Nuclear Information System (INIS)

    The irradiations which have been, and are being carried out in the Melusine and Siloe reactors in connection with pure or applied research projects, are effected in widely varying conditions; amongst these, for example, the temperature may vary from -250 deg C to +2000 deg C The eight devices presented are designed for irradiations effected at temperatures of from room temperature up to 2000 deg C. 1. Irradiation device for irradiation at normal temperatures 2. The 'PEF' device 3. The 'CHOUCA' device, 150 to 900 deg C 4. The 'CYRANO' device for EL 4 conditions 5. 'HT' capsules, 800-1000 deg C 6. The 'HEBE' furnace 1400 deg C 7. The 'PEC' device, 1400 deg C 8. The 'HF' furnace 2000 deg C. (authors)

  3. Preliminary conceptual design for electrical and I and C system of a new research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements.

  4. Preliminary conceptual design for electrical and I and C system of a new research reactor

    International Nuclear Information System (INIS)

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements

  5. Design and installation of DC plasma reactor for SiC nanoparticle production

    Science.gov (United States)

    Yu, I. K.; Rhee, J. H.; Cho, S.; Yoon, H. K.

    2009-04-01

    In order to get the high quality SiC nanopowders, a DC plasma reactor system with adjustable torch has been developed. SiC nanopowders were synthesized using this system and the synthesized primary particles have nearly spherical structures, mostly β-SiC phase with a particle size of 10-30 nm. Larger aggregate particles have been produced in our system probably due to longer particle growth times and faster collisions. The synthesized particles collected from the reactor wall and cyclone bottom have some free silicon and free carbon. To produce high quality silicon carbide nanopowders, it is highly necessary to improve the experimental conditions such as lower system pressures, shorter residence time, and higher quenching rates during powder synthesis.

  6. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  7. Creep crack growth in a reactor pressure vessel steel at 3600 C

    International Nuclear Information System (INIS)

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel

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

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

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

  9. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  10. Study on the chemical compatibility of SiC/SiC composites as core materials for Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Silicon carbide (SiC) fibers combined with a SiC matrix constitute an advanced solution for Fast Neutron Reactors (SFR, GFR…) as core materials for their stability at high temperature and their neutron transparency. However, the compatibility with the reactor's fuel and coolant has to be studied within a wide range of temperatures: moderated in normal operation (close-circuit at 550°C) and elevated (open-middle) in accidental conditions. Concerning the compatibility of SiC/SiC with the fuel and, considering in a first step uranium oxide, the stoichiometry of uranium dioxide remains one of the important parameters to be studied. Experimental results on the chemical compatibility between UO2.15 and SiC were obtained using high temperature mass spectrometry. The reaction for 6 hours at 1773 K produces the release of mainly COg, CO2(g) and SiO(g) and Si(g) and leads to the formation of uranium silicide compounds. The uranium oxide was fully consumed. Thermodynamic calculations were performed to prepare and interpret the experiments. (author). (author)

  11. Engineering and planning for reactor 105-C interim safe storage project subcontract no. 0100C-SC-G0001 conceptual design report. Volume 1

    International Nuclear Information System (INIS)

    The 105-C Reactor, one of eight surplus production reactors at the Hanford Site, has been proposed by the U.S. Department of Energy, Richland, Operations Office to be the first large-scale technology demonstration project in the decontamination and decommissioning (D ampersand D) focus area as part of the project for dismantlement and interim safe storage. The 105-C Reactor will be placed in an interim safe storage condition, then undergo the decontamination and decommissioning phase. After D ampersand D, the reactor will be placed in long- term safe storage. This report provides the conceptual design for these activities

  12. Evaluation of the 14C release characteristics of a reactor building using high-compression collection equipment in Korean pressurized water reactors

    International Nuclear Information System (INIS)

    Highlights: ► The emission from the reactor building is done in a batch-wise manner in a limited time. ► It is difficult to measure 14C with conventional gaseous 14C collection equipment due to the typically small sampling volume. ► Developed a device that can highly compress and store gas released over a short time in a tank. ► This device was successfully used to capture samples and assessed the characteristics of the 14C emissions from a reactor building. - Abstract: In general, most nuclear power plants (NPPs) have one exhaust duct. However, each Korean NPPs specifically equip four exhaust ducts that release gas into the environment. Thus, Korean NPPs have difficulty in managing the radioactive effluents. The gas from the reactor building at a nuclear power plant is released into the environment through an exhaust system while fresh air is supplied into the building at the same time to satisfy the access conditions of the reactor building. It is important to secure measurable samples while the emission is in progress for reliable measurements of 14C in the effluent gas from the reactor building and for reliable assessments of the radiation exposure dose to residents who live in the area. The emission from the reactor building is done in a batch-wise manner in a limited time. It is difficult to measure 14C with conventional gaseous 14C collection equipment due to the typically small sampling volume of this type of equipment. To address this problem, a device was developed that can highly compress and store gas released over a short time in a tank that can be connected to gaseous 14C collection equipment for later analysis. This device was successfully used to capture samples to assess the characteristics of the 14C emissions from a reactor building. In this paper, a high-compression sampling device developed for the analysis of 14C from a reactor building is introduced and the result of an assessment of the characteristics of 14C emission from the

  13. Numerical simulation research on C-ring of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    The seal element of CPR1000 reactor pressure vessel is C-ring , in this paper , five finite element models are established, including practical model, ring model, equivalent cylinder model. Using linear elastic analysis and elastoplastic analysis, each model's load-displacement curve of bearing radial load is obtained, which contains applying loading and unloading process. The result show that, equivalent cylinder model based on the medium diameter can reflect the characteristics of C-ring. At the same time, it can effectively reduce the cost of calculation, and can be used for seal analysis of reactor pressure vessel bolt-flange connecting structure. Analysis also indicates that, the alloy cladding and silver have a great influence on the results. (authors)

  14. Regulatory use the classification security systems of I and C in VVER type reactors

    International Nuclear Information System (INIS)

    Presently work the author proposes a classification to the system I and C to the VVER 440 type reactor in categories the regulatory control with a view to establishing the degree to the attention that the regulator should pay to these systems, leaving the importance that have the same ones for the security the installation, during the execution the works that are carried out with this equipment in the stages construction, setting in service and exploitation

  15. Final hazard classification and auditable safety analysis for the 105-C Reactor Interim Safe Storage Project

    International Nuclear Information System (INIS)

    This document summarizes the inventories of radioactive and hazardous materials present in the 105-C Reactor Facility and the operations associated with the Interim Safe Storage Project which includes decontamination and demolition and interim safe storage of the remaining facility. This document also establishes a final hazard classification and verifies that appropriate and adequate safety functions and controls are in place to reduce or mitigate the risk associated with those operations

  16. C-scope under-sodium viewer for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    A C-scope under-sodium viewer has been developed for monitoring the interior of sodium-cooled fast breeder reactors. Consisting of a transducer that emits and receives ultrasonic waves under liquid sodium, a mechanism that drives the transducer under liquid sodium and an image displaying section, it inspects the fuel assembly through its image in optically opaque high-temperature (3000C) liquid sodium. The results of its evaluation test are: (1) The transducer could continue satisfactory operation under 3500C (at the highest) sodium for more than a month. (2) The driving mechanism, though it was the first of the kind appearing in Japan, has been proved that it could continue operation for a week under 3000C sodium. (3) The image displaying section, in spite of the low speed of the transducer (below 20 rpm), could display stable and clear images. (4) The image in 3000C was as clear as that in room-temperature water. (auth.)

  17. Calibration of new I and C at VR-1 training reactor

    International Nuclear Information System (INIS)

    The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

  18. Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

    International Nuclear Information System (INIS)

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area

  19. Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.

    2000-02-11

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

  20. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B. [Washington State Univ., WA (United States); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Co., WA (United States)

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  1. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Diffusion of Fission Product Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H.; Jiang, Weilin

    2014-11-01

    MAX phases, such as titanium silicon carbide (Ti3SiC2), have a unique combination of both metallic and ceramic properties, which make them attractive for potential nuclear applications. Ti3SiC2 has been suggested in the literature as a possible fuel cladding material. Prior to the application, it is necessary to investigate diffusivities of fission products in the ternary compound at elevated temperatures. This study attempts to obtain relevant data and make an initial assessment for Ti3SiC2. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti3SiC2, SiC, and a dual-phase nanocomposite of Ti3SiC2/SiC synthesized at PNNL. Thermal annealing and in-situ Rutherford backscattering spectrometry (RBS) were employed to study the diffusivity of the various implanted species in the materials. In-situ RBS study of Ti3SiC2 implanted with Au ions at various temperatures was also performed. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti3SiC2 occurs during ion implantation at 873 K. Cs in Ti3SiC2 is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti3SiC2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Further studies of the related materials are recommended.

  2. A study on the sealing performance of metallic C-rings in reactor pressure vessel

    International Nuclear Information System (INIS)

    Highlights: • FE analysis on compression–resilience of metallic C-ring is performed and validated by experiments. • Model of RPV sealing system including the C-rings is developed. • Deformation data from factory hydraulic test of the RPV are used to verify the model. • C-rings’ behavior under designing condition is analyzed. • The model provides a reliable evaluation on the sealing performance of RPV. - Abstract: Double metallic C-rings are used in pressure vessel of pressurized water reactor (PWR) to seal the bolt-connected flanges. To evaluate the sealing performance, it is necessary to study both the C-rings’ intrinsic properties and their behavior in reactor pressure vessel (RPV) under various loading conditions. The compression–resilience property and linear load are the basic information to evaluate the performance of a well-designed C-ring's. An equivalent model of C-ring is constructed by means of ANSYS to analyze its intrinsic properties, and is also validated by experiments on scaled samples. This model is applied to develop a 2D-axisymmetric FE model of sealing system including RPV and C-rings with the consideration of nonlinear material, contacting problem and multiple coupled effects. The simulation results of RPV deformation under the hydraulic test condition agree well with the data of factory hydraulic test. With the verified model, an analysis under the designing condition is performed to study C-rings’ behavior in the RPV, and then provides a reliable evaluation on the sealing performance of RPV

  3. Design issues on using FPGA-based I and C systems in nuclear reactors

    International Nuclear Information System (INIS)

    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  4. Design issues on using FPGA-based I and C systems in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Santos, Isaac Jose A.L. dos; Lacerda, Fabio de, E-mail: msantana@ien.gov.br, E-mail: paulov@ien.gov.br, E-mail: luquetti@ien.gov.br, E-mail: acerda@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Engenharia Nuclear

    2015-07-01

    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  5. Physicochemical interactions resulting from the use of a SiC/SiC composite material in typical environments of future nuclear reactors

    International Nuclear Information System (INIS)

    The development of high purity SiC fibers during the nineties has led to their consideration as nuclear reactors components through the use of SiC/SiC composites. SiC and SiC/SiC composites are considered as core materials of future nuclear reactors (SFR, GFR) and as a potential replacement for the zirconium cladding of PWR. Therefore, the thermochemical compatibility of these materials with typical environments of those nuclear reactors has been studied. The composition and the growth kinetics of the reaction zone of SiC towards niobium and tantalum (considered as materials to ensure the leak-tightness of a SiC/SiC cladding for GFR) have been studied between 1050 and 1500 C. High temperature heat treatments in open and closed systems between SiC and UO2 have shown a significant reactivity over 1200 C characterized by the formation of CO and uranium silicides. Moreover, a liquid phase has been detected between 1500 and 1650 C. The exposure of SiC/SiC to liquid sodium (550 C, up to 2000 h) has been studied as a function of the oxygen concentration dissolved in liquid sodium. An improvement of the mechanical properties of the composites elaborated for this study (increase of the tensile strength and strain at failure) has been highlighted after immersion in the liquid sodium independently of its oxygen concentration. It is believed that this phenomenon is due to the presence of residual sodium in the material. (author)

  6. Model of a two-stage rf plasma reactor for SiC deposition

    Science.gov (United States)

    Petrov, G. M.; Giuliani, J. L.

    2001-07-01

    A reactor is proposed for plasma-enhanced chemical-vapor deposition of silicon carbide (SiC) at low pressure (˜few Torr). The inductively coupled plasma lies upstream of the growth substrate and serves to dissociate the precursor silane/propane/hydrogen inlet gas. Unlike existing reactors, the design offers the potential for separate control of the temperature in the dissociation region and at the growth substrate. The geometrical parameters and flow conditions appropriate for SiC growth are analyzed with a one-dimensional flow simulation model which includes approximations for lateral diffusive losses to cold walls as well as deposition to the substrate. Twenty-one neutral species and 24 ions are followed with 179 reactions. At 3 Torr, 10 W/cm3, and 300 cm/s inlet flow velocity, the model predicts a growth rate of ˜3 μm/h downstream from the plasma. Negligible ion density exists over the substrate as long as the silane density is sufficiently large due to a feedback process between Si+ and SiH4. Besides heating the gas, the plasma is an efficient source of radical H atoms, which in turn control the abundance of some hydrocarbon species over the substrate. C2H2 is the dominant contributor to the C-bearing flux onto the substrate and the Si atom, which forms by electron reactions, is the most important Si-bearing species. Finally, a sensitive transition in deposition rate is found for the C-bearing species as the power increases from 5 to 10 W/cm3.

  7. Eu2O3 and B4C worth calculations in fast reactor spectra

    International Nuclear Information System (INIS)

    Work is now in progress to design and fabricate europia (Eu2O3) control rods for irradiation testing in one of the early operating cycles of the Fast Test Reactor (FTR). These tests will provide data for evaluating europia as a possible control rod absorber material in fast reactors. Prediction of the reactivity worth of the europia rods is a necessary part of the design process. To improve the accuracy of these predictions, an experiment was performed in the FTR Engineering Mockup Critical (EMC) to determine the relative reactivity worth of Eu2O3 and boron carbide (B4C) in control rod size quantities. Upon completion of the experiment, calculations were performed to obtain reactivity worths for comparison with the measured worths. From this comparison, calculation-vs.-experiment (C/E) bias factors were obtained for use in correcting the computed reactivity worth of europia in the FTR. A brief description of the experiment is presented together with the experimental results, a description of the analytical methods, the calculated results, and a comparison of the calculated and experimental results. Also included are discussions of the effects of resonance self-shielding and mesh spacing on computed absorber rod worths

  8. Methane Conversion to C2 Hydrocarbons in Solid State Oxide Electrolyte Membrane Reactor

    Institute of Scientific and Technical Information of China (English)

    LI Jun; ZHAO Ling; ZHU Zhong-nan; XI Dan-li

    2005-01-01

    Provskite-type catalysts, Ln0.6 Sr0.4 FexCo1-x O3 (Ln = Nd,Pr, Gd, Sm, La, 0<x<1) and Ln0.8Na0.2CoO3(Ln= La,Gd, Sm) were synthesized, their catalytic properties in the oxidative coupling of methane (OCM) were examined in a fixed-bed reactor. The former group presented higher activity in the OCM, but the main product was carbon dioxide. While the later group showed lower activity but much higher selectivity to C2 hydrocarbons compared with the former. Electrochemical measurements were conducted in a solid oxide membrane reactor with La0.8 Na0.2CoO3 as catalyst. The results showed that methane was oxidized to carbon dioxide and ethane by two parallel reactions. Ethane was oxidized to ethene and carbon dioxide. A fraction of ethene was oxidized deeply to carbon dioxide. The total selectivity to C2 hydrocarbons exceeded 70%. Based on the experimental results, a kinetic model was suggested to describe the reaction results.

  9. Biohydrogen production from glucose in upflow biofilm reactors with plastic carriers under extreme thermophilic conditions (70(degree)C)

    DEFF Research Database (Denmark)

    Zheng, H.; Zeng, Raymond Jianxiong; Angelidaki, Irini

    2008-01-01

    Biohydrogen could efficiently be produced in glucose-fed biofilm reactors filled with plastic carriers and operated at 70°C. Batch experiments were, in addition, conducted to enrich and cultivate glucose-fed extremethermophilic hydrogen producing microorganisms from a biohydrogen CSTR reactor fed...... with synthetic medium with glucose as the only carbon and energy source. A biofilm reactor, started up with plastic carriers, that were previously inoculated with the enrichment cultures, resulted in higher hydrogen yield (2.21 mol H2/mol glucose consumed) but required longer start up time (1 month), while...

  10. Design and Testing of D.C. Conduction Pump for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    DC Conduction pump immersed in sodium forms a part of Failed Fuel Location Module (FFLM) of 500 MWe Fast Breeder Reactor (PFBR) currently under construction. FFLM housed in control plug of the reactor, is used to locate the failed fuel sub-assembly due to clad rupture in the fuel pin. The DC conduction pump sucks the sodium from the top of fuel sub-assemblies through the selector valve and pumps the sodium to hold up for detecting the presence of delayed neutrons. Presence of delayed neutron is the indication of failure in the sampled fuel sub-assembly. The DC Conduction Pump was chosen because of its low voltage operation (2 V) where argon/alumina ceramic can provide required electrical insulation even at operating temperature of 560 deg. C without much complication on the manufacturing front. Sampling of sodium from top of different sub-assemblies is achieved by operation of selector valve in-conjunction with the drive motor. FFLM requires the pump to be immersed in sodium pool at ∼ 560 deg. C located above the fuel sub-assemblies in the reactor. The Pump of 0.36 m3/h capacity and developing 1.45 Kg/ cm2 pressure was designed, manufactured and tested. The DC Conduction Pump has a stainless steel duct filled with liquid sodium, which is to be pumped. The stainless steel duct is kept in magnetic field obtained by means of electromagnet. The electromagnet is made of soft iron and the coil made of copper conductor surrounds the yoke portion of electromagnet. The external DC source of 2000 Amps, 2 Volt is used to send current through sodium placed in the stainless steel duct and the same current is sent through copper coil of electromagnet for producing required magneto motive force, which in turn produces required magnetic field. The interaction of current in sodium (placed in stainless steel duct) and magnetic field produced by the electromagnet in the duct region produces pumping force in the sodium. Electromagnet, copper coil, stainless steel duct, copper bus

  11. Thermophilic (55 - 65°C) and extreme thermophilic (70 - 80°C) sulfate reduction in methanol and formate-fed UASB reactors

    NARCIS (Netherlands)

    Vallero, M.V.G.; Camarero, E.; Lettinga, G.; Lens, P.N.L.

    2004-01-01

    The feasibility of thermophilic (55-65 degreesC) and extreme thermophilic (70-80 degreesC) sulfate-reducing processes was investigated in three lab-scale upflow anaerobic sludge bed (UASB) reactors fed with either methanol or formate as the sole substrates and inoculated with mesophilic granular slu

  12. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored

  13. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  14. A Behavior-Preserving Translation From FBD Design to C Implementation for Reactor Protection System Software

    International Nuclear Information System (INIS)

    Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation

  15. A Behavior-Preserving Translation From FBD Design to C Implementation for Reactor Protection System Software

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Junbeom; Kim, Euisub [Konkuk Univ., Seoul (Korea, Republic of); Lee, Jangsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-08-15

    Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation.

  16. Evaluation of the neutrons spectrum near the Venus reactor: use of MCNPX-2.5C

    International Nuclear Information System (INIS)

    The present study has been justified by the choice of the Venus reactor (SCK-CEN) as a true work environment for the project of the fifth programme - frame E.V.I.D.O.S.. The objective of this programme is the evaluation, in neutron-photon combined field, and in true environment (nuclear industry), of the different methods of measurement used in neutron dosimetry. The project aims to the determination of abilities and limits of dosemeters and to establish methods to get doses equivalents from data gotten by spectrometry, personal and ambient dosimetry. For each environment, reference values have to be determined by spectrometry (energy and angle). The knowledge of the distribution in energy and in angle of neutrons allows then the calculation of the different doses equivalents. The determination of these references values by direct neutron calculation allows the validation of the Monte Carlo model. (N.C.)

  17. MCNPX 2.6.C versus MCNPX and ORIGEN-S: state of the art for reactor core management

    International Nuclear Information System (INIS)

    This paper discusses the application of the Monte Carlo burnup code MCNPX-2.6.C for the criticality and depletion reactor core analysis of the Material Testing Research Reactor BR2 in Mol, Belgium. A comparison with the combined MCNP and ORIGEN-S fuel depletion method is presented. The accuracy of both methods, the consumption of calculation time, the depletion capabilities, the advantages and disadvantages of use of both methods are discussed. The accuracy of the criticality calculations by MCNPX-2.6.C is still lower in comparison with the validated MCNP and ORIGEN-S method

  18. Joining of SiC/SiCf ceramic matrix composites for fusion reactor blanket applications

    International Nuclear Information System (INIS)

    Using a preceramic polymer, joints between SiC/SiCf ceramic matrix composites were obtained. The polymer, upon pyrolysis at high temperature, transforms into a ceramic material and develops an adhesive bonding with the composite. The surface morphology of 2D and 3D SiC/SiCf composites did not allow satisfactory results to be obtained by a simple application of the method initially developed for monolithic SiC bodies, which employed the use of a pure silicone resin. Thus, active or inert fillers were mixed with the preceramic polymer, in order to reduce its volumetric shrinkage which occurs during pyrolysis. In particular, the joints realized using the silicone resin with Al-Si powder as reactive additive displayed remarkable shear strength (31.6 MPa maximum). Large standard deviation for the shear strength has nevertheless been measured. The proposed joining method is promising for the realization of fusion reactor blanket structures, even if presently the measured strength values are not fully satisfactory

  19. Recovery of 14C from graphite moderator of gas-cooled reactor (GCR)

    International Nuclear Information System (INIS)

    The chemical exchange method of carbon isotopes between CO2 and carbamate was applied to the recovery of 14C from 1,600 t graphite moderator of a gas-cooled reactor (GCR), Tokai-1, and the dimensions of 14C-enrichment process were evaluated numerically. Applicability of two processes with different operation modes, continuous process and batch process, was discussed under the conditions that the concentration of 14CO2 in the stripped flow corresponding to 99% of feed CO2 is less than the environmental standard. For the continuous process using 2 mol/l diethylamine (DEA)-octane solution as a working fluid at -20degC and 0.2 MPa, the column dimensions were evaluated as 3.2 m in diameter and 5.7 m in height in the case of operating period of 20 yr. For the batch process using 4 mol/l DEA-octane solution, the column dimensions were comparable to those of continuos process, when the process was operated at the rate of 4 batch/month under the conditions of -20degC and 0.3 MPa. From these results, it is concluded that the CO2/carbamate exchange method is applicable to the recovery of 14C from irradiated graphite. However, the batch process has serious disadvantages, such as large energy consumption to maintain the top reservoir at low temperature and the generation of a large quantity of secondary wastes. At the present stage, the continuous process should be selected for the practical process design. (author)

  20. Transportable nuclear power plant T3C-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Development and cultivating of districts in Siberia, North, Far East, Kamchatka and other remote or almost inaccessible district of the country depends to a large degree on their providing with power. The specific character of these districts imposes in turn a wide variety of special requirements upon the power sources. In particular, it is essential to provide the following; maximum manufacture availability of the whole equipment at the minimum volume of construction and installation work on operation site, high safety, longterm service life, ecologically, minimum scope of work on equipment in-service maintenance and inspection, etc. Taking into account the well-known difficulties connected with the delivery of conventional energy carriers to the above-mentioned districts and the situation with the alternative power sources, the application of the low-power nuclear plants (NPP) for these purposes looks definitely promising. Among the probable trends in creating the NPPs of this type as very promising is considered the possibility to apply the two-circuit reactor plant of the vessel type with the liquid lead as a primary coolant and free air as a secondary coolant and working medium in the open gas-turbine cycle. The nuclear plant T3C-M of improved safety with two of this type reactor plants with total electric power of 8 MW is developed by CDB of Machine Building with participation of several enterprises of St. Petersburg under the scientific leadership and is intended for generation of electric power and up to 4 Gcal/h of heat for populated areas and installations placed at long distance from the main electric power supply sources where it is difficult or non-efficient economically to deliver the conventional kinds of fuel. The main principles being laid as a basis when developing the proposed NPP will allow one to create mobile power sources which possess a high degree of safety and inherent self-protection

  1. First experiences from system integration, installation and commissioning of TELEPERM XS for reactor I and C at the Unterweser NPP

    International Nuclear Information System (INIS)

    The modernization of Reactor I and C, consisting of reactor limitation system, reactor control system and rod control system, at Unterweser NPP is the pilot application of the state-of-the-art safety I and C system TELEPERM XS. The Unterweser system has been integrated and tested from December 1996 to May 1997 in the Siemens Erlangen test field and has been installed at site in July 1997. For the period from July 1997 to Jul 1998 the new TELEPERM XS based Reactor I and C system will be operated online-open-loop in parallel to the existing system, in order to get information about the long term stability of the system and conduct intensive personnel training. For one selected function ''Power distribution control'' the operator has the possibility to choose between the old controller and the new TELEPERM XS function. During the 1998 outage the TELEPERM XS system will be connected to the process and the old I and C system will be dismantled. This document describes the experiences gathered during system integration in the test field. (author)

  2. Effect of NaCl on thermophilic (55°C) methanol degradation in sulfate reducing granular sludge reactors

    NARCIS (Netherlands)

    Vallero, M.V.G.; Hulshoff Pol, L.W.; Lettinga, G.; Lens, P.N.L.

    2003-01-01

    The effect of NaCl on thermophilic (55degreesC) methanol conversion in the presence of excess of sulfate (COD/SO42-=0.5) was investigated in two 6.5L lab-scale upflow anaerobic sludge bed reactors inoculated with granular sludge previously not adapted to NaCl
    The effect of NaCl on thermophilic (

  3. Fault tolerant distributed real time computer systems for I and C of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Architecture of distributed real time computer system (DRTCS) used in I and C of PFBR is explained. • Fault tolerant (hot standby) architecture, fault detection and switch over are detailed. • Scaled down model was used to study functional and performance requirements of DRTCS. • Quality of service parameters for scaled down model was critically studied. - Abstract: Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation and control (I and C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I and C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network

  4. Core design and safety analyses of 600 MWt, 950 °C high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Masaaki, E-mail: nakano-m@fujielectric.co.jp [Fuji Electric Co., Ltd., 1-1, Tanabe-shinden, Kawasaki-ku, Kawasaki-city 210-9530 (Japan); Takada, Eiji; Tsuji, Nobumasa; Tokuhara, Kazumi; Ohashi, Kazutaka; Okamoto, Futoshi [Fuji Electric Co., Ltd., 1-1, Tanabe-shinden, Kawasaki-ku, Kawasaki-city 210-9530 (Japan); Tazawa, Yujiro; Tachibana, Yukio [Japan Atomic Energy Agency, Oarai, Ibaraki-pref. 311-1393 (Japan)

    2014-05-01

    The conceptual core design study of high temperature gas-cooled reactor (HTGR) is performed. The major specifications are 600 MW thermal output, 950 °C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, Ag-110m and Cs-137 from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.

  5. Optical and electrical study of Ar-C2 H2 and N2-C2 H2 RF-excited plasmas in a small cylindrical reactor

    International Nuclear Information System (INIS)

    An optical and electrical investigation were carried out of organic plasmas obtained from DC and RF discharges in mixtures of Arc-C2 H2 in a small cylindrical reactor operating under pressures ranging from 0.3 to 1.0 Torr. The Actinometric Optical Emission Spectroscopy was used to follow the trends of C H and H species in the reactive plasma, for different pressures and power coupled to the reactor. The results showed a decreasing of C H concentration with the increasing of pressure and DC power and an increasing of the H concentration with the increasing of the pressure. The electrical investigation was performed using pulsed Langmuir probe technique, indicating typical electronic temperature around 2 eV. The effects of film deposition on the probe tip surface was analyzed using. Ar as reference gas. The results obtained indicated application of the probe during few period of sweep cycle. (author). 8 refs, 5 figs

  6. Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

    International Nuclear Information System (INIS)

    Prior work has demonstrated the use of a natural B4C capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method to critical assemblies for performing nuclear data measurements at near 235U fission-energy neutron spectrum. Previous fission product measurements showed that the neutron spectrum achievable with natural B4C was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a B4C capsule 96 % enriched in 10B resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. Fission product yields measured following an irradiation of a sample with this new method and subsequent radiochemical separations are presented here. (author)

  7. Fuel of plutonium monocarbide and inert diluent solid solution (54.5% PuC + 45.5% ZrC) for fast reactors

    International Nuclear Information System (INIS)

    The problem of efficient and cost-effective use of plutonium without breeding in nuclear power is the urgent scientific and engineering task. Solution of this problem in practice is related to the development of closed fuel cycle and utilization of fast reactors, including Pu and minor actinides burner fast reactors. Use of plutonium fuel with inert diluent that substitutes 238U and eliminates plutonium breeding is the problem of great practical interest. All properties of solid solutions and composite materials of carbides and nitrides (PuC - ZrC, PuN - ZrN, PuN - AIN etc.) comply with the requirements imposed on nuclear fuel to the greatest extent. Efficient plutonium burning is aided by increased plutonium content in the fuel. ARSRIIM was the first to propose the fuel based on PuC - ZrC and PuN - ZrN solid solutions with high thermal conductivity and thermodynamic stability that is now under development. Methods for synthesizing solid solutions of plutonium and zirconium monocarbides through entire range of compositions was developed by the authors, the procedure to produce 54.5% PuC + 45.5% ZrC fuel was chosen, irradiation test was carried out in BOR-60 reactor with 400 - 500 W/cm heat rating up to 8.0% HM burnup. As a result of irradiation tests it was concluded that all fuel elements remained their serviceability and were suitable for further operation, fuel-cladding interaction was local and did not exceed 100 μm, fuel swelling was about 1%/% of burnup, gas release was less than 2%. It was for the first time, when the possibility to use fuel with high content of plutonium and inert diluent in fast reactor was shown. (author). 7 refs, 1 fig., 3 tabs

  8. Improvement of Biohydrogen Production under Increased the Reactor Size by C. acetobutylicum NCIMB 13357

    Directory of Open Access Journals (Sweden)

    Hisham S. Alshiyab

    2009-01-01

    Full Text Available Problem statement: One of the main factors influenced the bacterial productivity and total yield of hydrogen is the partial pressure of produced gas. A novel solution to enhance the bacterial productivity was through reduction of gas pressure. Approach: Increasing the reactor size showed to enhance the bacterial production of hydrogen. Results: The technique of increasing reactor size resulted to enhance the hydrogen yield (YP/S from 269 mL g-1 glucose utilized to maximum yield of 448 mL g-1 glucose utilized by using 125 mL and 2 L reactor size respectively. The hydrogen productivity was also enhanced from 71 mL-1 h-1 to maximum of 91 mL L-1 h-1 was obtained by using 125 mL and 1 L reactor size respectively. Biomass concentration was enhanced from 1.03 g L-1 to maximum of 1.68 g L-1 by using 125 mL and 2 L reactor size were used respectively, hydrogen yield per biomass (YP/X of 267 mL g-1 L-1, biomass per substrate utilized (YX/S of 0.336 and produced hydrogen in gram per gram of glucose utilized (YH2/s of 0.04 when 2 L reactor size was employed. Conclusion: By using bigger reactor size, the effect of gaseous products in fermentation medium was reduced and enhanced both bacterial productivity and biomass concentration.

  9. Design and construction of a demineralizer for primary water of N.R.C. reactor

    International Nuclear Information System (INIS)

    One of the most important problems facing pool type research reactors is purification of its primary coolant from the minerals such as Ca, Al and Mg which are normally found in raw water of fission products produced due to nuclear reactions in reactor core. If these minerals are present in the reactor pool (primary coolant) they will be activated and the activity of primary coolant will go up. Therefore, a water purification device called demineralizer is needed to keep the pool water resistivity at 106 OHM/CC at all times. The reactor at Nuclear Research Center has a demineralizer with a flow of 15 G.P.M. that is unable to purify the reactor primary coolant from activated minerals. Therefore, the reactor operating group decided to design and manufacture a system with a higher capacity of purification with the cooperation of the research center workshops and to install it in the reactor primary coolant, so that the water in the primary coolant be circulated with a flow of 60 G.P.M. constantly. This is quite important since the primary coolant which is used for cooling, moderating shielding and reflection must be completely free of any minerals, otherwise the minerals in the primary coolant are activated due to reactions with neutrons in the core and consequently the activity inside the reactor hall will go up. Therefore a demineralizer with a higher capacity to remove the minerals seems to be necessary. Benefits: 1. The manufacture of a demineralizer and learning about its production technique is a step toward self-sufficiency which could be used in all factories and installations utilizing make up water. 2. Preparation of reactor in order to improve its function and lengthening its useful lifetime. 3. Utilization of this project for education of chemistry and mechanical students are among the goals of this project

  10. In situ reactor radiation-induced attenuation in sapphire optical fibers heated up to 1000 °C

    International Nuclear Information System (INIS)

    The purpose of this work was to determine the suitability of using instrumentation utilizing sapphire optical fibers in a high temperature nuclear reactor environment. For this, the broadband (500–2200 nm, or 0.56–2.48 eV) optical transmission in commercially available sapphire optical fibers was monitored in situ prior to, during, and after reactor irradiation. Portions of the sapphire fibers were heated to temperatures up to 1000 °C during irradiation. The sapphire fibers were irradiated, mostly at a neutron flux of 5.0 × 1011 n/cm2/s and a gamma dose rate of 28 kGy/h (dose in sapphire), to a total neutron fluence of 6.4 × 1016 n/cm2 and total gamma dose on the order of 1 MGy. Results were generally consistent with the results of previous in situ measurements of the transmission in unheated sapphire fibers during reactor irradiation. Added attenuation at 850, 1300, and 1550 nm, appears to be limited by the growth of radiation-induced defect centers that are located in the ultra violet to the visible range and is therefore less at 1300 and 1550 nm than at 850 nm. A linear increase in attenuation, due to displacement damage effects, was observed with increased irradiation time at constant reactor power. However, the rate of increase of the added attenuation during constant power reactor irradiation monotonically decreased with increasing temperature up to 1000 °C, with the most significant decrease occurring between 300 and 600 °C. Additional calculations predicted that the majority of (if not all of) the observed increases in attenuation during irradiation at 600 and 1000 °C were due to effects in the unheated sections of the irradiated sapphire fibers. These results suggest that, for a reactor radiation environment similar to that tested in this work, heating sapphire fibers to temperatures of 600 °C or greater during irradiation would significantly reduce (or possibly eliminate entirely) the rate of growth of the added attenuation in the sapphire

  11. Application of macro-cellular SiC reactor to diesel engine-like injection and combustion conditions

    Science.gov (United States)

    Cypris, Weclas, M.; Greil, P.; Schlier, L. M.; Travitzky, N.; Zhang, W.

    2012-05-01

    One of novel combustion technologies for low emissions and highly efficient internal combustion engines is combustion in porous reactors (PM). The heat release process inside combustion reactor is homogeneous and flameless resulting in a nearly zero emissions level. Such combustion process, however is non-stationary, is performed under high pressure with requirement of mixture formation directly inside the combustion reactor (high pressure fuel injection). Reactor heat capacity resulting in lowering of combustion temperature as well as internal heat recuperation during the engine cycle changes the thermodynamic conditions of the process as compared to conventional engine. For the present investigations a macro-cellular lattice structure based on silicon carbide (non-foam structure) with 600 vertical cylindrical struts was fabricated and applied to engine-like combustion conditions (combustion chamber). The lattice design with a high porosity > 80% was shaped by indirect three-dimensional printing of a SiC powder mixed with a dextrin binder which also serves as a carbon precursor. In order to perform detailed investigations on low-and high-temperature oxidation processes in porous reactors under engine-like conditions, a special combustion chamber has been built and equipped with a Diesel common-rail injection system. This system simulates the thermodynamic conditions at the time instance of injection onset (corresponding to the nearly TDC of compression in a real engine). Overall analysis of oxidation processes (for variable initial pressure, temperature and air excess ratio) for free Diesel spray combustion and for combustion in porous reactor allows selection of three regions representing different characteristics of the oxidation process represented by a single-step and multi-step reactions Another characteristic feature of investigated processes is reaction delay time. There are five characteristic regions to be selected according to the delay time (t) duration

  12. Effect of the C:N:P ratio on the denitrifying dephosphatation in a sequencing batch biofilm reactor (SBBR).

    Science.gov (United States)

    Mielcarek, Artur; Rodziewicz, Joanna; Janczukowicz, Wojciech; Thornton, Arthur J; Jóźwiak, Tomasz; Szymczyk, Paula

    2015-12-01

    A series of investigations were conducted using sequencing batch biofilm reactor (SBBR) to explore the influence of C:N:P ratio on biological dephosphatation including the denitrifying dephosphatation and the denitrification process. Biomass in the reactor occurred mainly in the form of a biofilm attached to completely submerged disks. Acetic acid was used as the source of organic carbon. C:N:P ratios have had a significant effect on the profiles of phosphate release and phosphate uptake and nitrogen removal. The highest rates of phosphate release and phosphate uptake were recorded at the C:N:P ratio of 140:70:7. The C:N ratio of 2.5:1 ensured complete denitrification. The highest rate of denitrification was achieved at the C:N:P ratio of 140:35:7. The increase of nitrogen load caused an increase in phosphates removal until a ratio C:N:P of 140:140:7. Bacteria of the biofilm exposed to alternate conditions of mixing and aeration exhibited enhanced intracellular accumulation of polyphosphates. Also, the structure of the biofilm encouraged anaerobic-aerobic as well as anoxic-anaerobic and absolutely anaerobic conditions in a SBBR. These heterogeneous conditions in the presence of nitrates may be a significant factor determining the promotion of denitrifying polyphosphate accumulating organism (DNPAO) development. PMID:26702975

  13. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    International Nuclear Information System (INIS)

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A ampersand 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met

  14. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  15. Regulatory involvement in IandC systems upgrading on WWER 440 type reactors in the Slovak Republic

    International Nuclear Information System (INIS)

    An overview is given of the recommendations and regulations concerning IandC systems of nuclear power plants with WWER-440 reactors in the Slovak Republic, and of the relevant involvement of regulatory bodies. The issues included in regulatory decrees and pertaining to the safety aspects of IandC systems are mentioned point by point, with emphasis put on the upgrading of the systems. The power plants covered include the Bohunice V-1 and V-2 plants and the Mochovce-1 and Mochovce-2 units. (A.K.)

  16. Greater-than-Class C low-level waste characterization. Appendix F: Greater-than-Class C low-level radioactive waste light water reactor projections

    International Nuclear Information System (INIS)

    This study characterizes potential greater-than-Class C low-level radioactive waste streams, estimates the amounts of waste generated, and estimates their radionuclide content and distribution. Several types of low-level radioactive wastes produced by light water reactors were identified in an earlier study as being potential greater-than-Class C low-level waste, including specific activated metal components and certain process wastes in the form of cartridge filters and decontamination resins. Light water reactor operating parameters and current management practices at operating plants were reviewed and used to estimate the amounts of potential greater-than-Class C low-level waste generated per fuel cycle. The amounts of routinely generated activated metal components and process waste were estimated as a function of fuel cycle. Component-specific radionuclide content and distribution was calculated for activated metals components. Empirical data from actual low-level radioactive waste streams were used to estimate radionuclide content and distribution for process wastes. The greater-than-Class C low-level waste volumes that could be generated through plant closure were also estimated, along with volumes and activities for potential greater-than-Class C activated metals generated at decommissioning

  17. How to improve the uniformity of deposition conditions in large pulsed D.C. PACVD reactors

    International Nuclear Information System (INIS)

    The deposition of hard coatings on different types of work-pieces by plasma-assisted chemical vapour deposition (PACVD) has become an important process in industry. In order to increase the productivity of this technique there was a demand in an up-scaling. However, a few problems are related with this task. Since a couple of years it is known, that a gas distribution system is required for a uniform thickness distribution and quality of the deposited films. However, it became evident only very recently that the dynamics of the discharge has a significant influence on the homogeneity of the coatings as well. Investigations of the temporal and spatial evolution of unipolar pulsed direct-current (d.c.) discharges relevant for the production of titanium nitride (TiN) coatings using titanium tetrachloride (TiCl4) as a feedstock gas have shown that under certain loading conditions the spreading of the discharge during the pulse is too slow. The discharge ignites at the beginning of each pulse at one spot and spreads from there across the reactor. The ignition and spreading behaviour of the discharge can be very much improved by the superposition of short high voltage pulses at the beginning of each conventional pulse. These additional short high voltage pulses produce additional charge carriers at the beginning of the conventional negative pulses and enable a fast ignition and spreading. Another possibility to improve the ignition and spreading behaviour of the discharge is the use of bipolar voltage pulses. Some commercially available power supplies have already included this feature. During the positive voltage pulses the ratio of the surfaces of cathode to anode is inverted compared to the case of the negative voltage pulse. Because of the large surface of the cathode during the positive voltage pulse the discharge current and accordingly the plasma power is significantly higher. So one can assume that at the end of the pulse-off time more charge carriers remain and

  18. Steady state thermal-hydraulic analysis using computer code cobra-3c/rertr for low power research reactors

    International Nuclear Information System (INIS)

    A number of codes are available for performing the steady state and transient thermal-hydraulics of nuclear reactors Viz RELAP,COBRA,PARET,etc. However COBRA-3C/RERTR is a verstile tool for the thermal hydraulic analysis of research reactors with steady state and transient capability . The steady state capability of the code is tested for a reference core of research reactor similar to PARR. Reference core under consideration is of 6x4 configuration with UAL-Al fuel and with 18 standard and 6 control elements. Safety margins against different critical phenomena like onset of nucleate boiling, departure from nucleate boiling and onset of flow instability are evaluated as a function of operating power and flow rate. Results indicate that COBRA-3C/RERTR can be used for a range of steady state parametric evaluations of PARR-1 type cores with any type of fuel used i.e. LEU, HEU, silicide, Oxide or aluminide with plate type geometry. The results reported, demonstrate the useful capability of the code and may provide a basis for comparison of results for any type of LEU, HEU or mixed cores, in future

  19. COMMIX-1C code estimation for the pool dynamics of Istanbul Technical University TRIGA MARK-II reactor

    International Nuclear Information System (INIS)

    In this study, the COMMIX-1C code is used to investigate the pool dynamics of Istanbul Technical University (ITU)TRIGA MARK-II reactor by simulating the velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. COMMIX-1C is multi-purpose, three-dimensional. transient, single-phase, thermal-hydraulics computer code. For the mass, momentum and energy equations, it uses a porous-medium formulation, a finite-volume algorithm, a flow modulated skew-upwind discretization scheme to reduce numerical diffusion and k-ε two-equation turbulence model. Its implementation for the particular system requires geometric and physical modelling decisions. ITU TRIGA MARK-II reactor pool is considered partly as continuum and partly as porous medium. All the major pool components are explicitly modelled in the simulation. Shape of the pool structure and computational cells are accounted for using the concept of directional surface permeability, volume porosity, distributed resistance, and distributed heat source or sink. The results are compared to the results of the computer codes TRISTAN, TRIGATH and TRIGATH-R

  20. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [pnnl; Alvine, Kyle J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roosendaal, Timothy J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shin, Yongsoon [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nguyen, Ba Nghiep; Borlaug, Brennan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jiang, Weilin [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Arreguin, Shelly A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-01-15

    A new dual-phase nanocomposite of Ti₃SiC₂/SiC is being synthesized using preceramic polymers, ceramic powders, and carbon nanotubes (CNTs) designed to be suitable for advanced nuclear reactors and perhaps as fuel cladding. The material is being designed to have superior fracture toughness compared to SiC, adequate thermal conductivity, and higher density than SiC/SiC composites. This annual report summarizes the progress towards this goal and reports progress in understanding certain aspects of the material behavior but some shortcomings in achieving full density or in achieving adequate incorporation of CNTs. The measured thermal conductivity is adequate and falls into an expected range based on SiC and Ti₃SiC₂. Part of this study makes an initial assessment for Ti₃SiC₂ as a barrier to fission product transport. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti₃SiC₂, SiC, and a synthesized at PNNL. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti₃SiC₂ occurs during ion implantation at 873 K. Cs in Ti₃SiC₂ is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti₃SiC₂ as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Progress is reported in thermal conductivity modeling of SiC-based materials that is relevant to this research, as is progress in modeling the effects of CNTs on fracture strength of SiC-based materials.

  1. Post-irradiation examination of mixed (Pu, U)C fuels irradiated in the fast breeder reactor

    International Nuclear Information System (INIS)

    The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India, using mixed (Pu,U)C fuel has completed eleven years of operation. One fuel subassembly which has seen more than 25,000 MWd/t burn-up has been discharged from this reactor taken up for post-irradiation examination. The PIE carried out on this fuel subassembly has established that the fuel has performed satisfactorily and it is capable of being taken to higher levels of burn-up and linear heat ratings. The facilities available for PIE of advanced fuels and the PIE work carried out are discussed in detail in this paper. (author)

  2. Operation and maintenance experiences at the C.R.E. Casaccia TRIGA reactor

    International Nuclear Information System (INIS)

    The memoir explains TRIGA RC-1 plant activities from last European TRIGA Users' Conference till today. In particular, measures following reactor exercise license renewing (March 1987) are described. Finally, difficulties and measures about shielding tank's water funguses and spores contamination, are explained. (author)

  3. Design criteria -- Reactor plant modifications for increased production and 100-C Area Alterations (Sections A and B) CG-558. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Russ, M.H.

    1954-08-10

    This document defines the basic criteria to be used in the preparation of detailed design for Project CG-558, Reactor Plant Modification for Increased Production and for Project CG-600, 100-C Area Alterations. It has been determined that the most economical method of increasing plutonium production within the next five years is by the modernization and improvement of the 100-B, 100-C, 100-D, 100-DR, 100-F, and 100-H reactor plants. These reactors are currently incapable of operating at their maximum potential power levels because of a limited availability of process cooling water. As a result of this programs, it is estimated that 1650-2350 megawatts of total additional production will be achieved. The purpose of this document is to set forth the design for certain modifications and additions to Hanford reactors and their supporting facilities as required to obtain higher power levels and improve the safety of reactor operation.

  4. Horizontal hot wall reactor design for epi-SiC growth

    Energy Technology Data Exchange (ETDEWEB)

    Veneroni, A.; Omarini, F.; Masi, M. [Dipartimento di. Chimica Materiali e Ingegneria Chimica ' ' Giulio Natta' ' , Politecnico di Milano via Mancinelli 7, 20131 Milano (Italy); Leone, S.; Mauceri, M.; Pistone, G.; Abbondanza, G. [Epitaxial Technology Center, Contrada Torre Allegra, 95121Catania (Italy)

    2005-11-01

    The model adopted for the simulation of a new industrial size type of horizontal cold wall reactor for epitaxial silicon carbide deposition is reviewed. The attention is focalized on the chemical mechanism adopted and on the comparison with some growth rate data and temperature profiles for the system ethylene, silane, hydrogen and the deposition of undoped silicon carbide. (copyright 2005 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  5. The C language auto-generation of reactor trip logic caused by steam generator water level using CASE tools

    International Nuclear Information System (INIS)

    The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsung 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using statechart-based formalism and statemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language through we manually made Software Requirement specification(SRS) for safety-critical software using statechart-based formalism. Most of the phase of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering(CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation. (Author). 6 refs., 6 figs

  6. Current Zero Missing Phenomena caused by D.C. Current which Flows from Shunt Reactor at the Ground Fault and its Interruption

    Science.gov (United States)

    Kobayashi, Takayuki; Eto, Atsushi; Koshizuka, Tadashi; Nishiwaki, Susumu; Kudo, Kietsu; Matsushita, Kozo; Hosokawa, Osamu

    At substations where shunt reactors are installed, when a ground fault occurs near the reactor, d.c. current flows from the reactor to the fault point. There are cases where circuit breakers installed near the reactor interrupt only the d.c. current from the reactor with long arcing times. In those cases, attention should be paid to the damage of the contacts. This paper shows the investigation of the d.c. current interrupting performance for the 550kV one-break SF6 gas circuit breaker. In the range of some hundreds amperes, the d.c. arc voltages with long arcing times of the 550kV one-break circuit breaker were measured. The d.c. current waveform and arcing time at the d.c. current interruption by 550kV one-break circuit breaker in the 500kV cable system were calculated. Interrupting test circuit for the d.c. current interruption was constructed in the high power laboratory. From the interrupting test results, it was made clear that the circuit breaker had a good interrupting performance after the d.c. current interruptions.

  7. Removal Site Evaluation Report to the C-Reactor Seepage Basins (904-066, -067 and -068G)

    International Nuclear Information System (INIS)

    Removal Site Evaluation Reports are prepared in accordance with Section 300.410 of the National Contingency Plan (NCP) and Section X of the Federal Facility Agreement (FFA). The C-Reactor Seepage Basins (904-066G,-067G,-068G) are listed in Appendix C, Resource Conservation and Recovery Act (RCRA)/Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Units List, of the FFA. The purpose of this investigation is to report information concerning conditions at this unit sufficient to assess the threat (if any) posed to human health and the environment and to determine the need for additional CERCLA action. The scope of the investigation included a review of past survey and investigation data, the files, and a visit to the unit.Through this investigation unacceptable conditions of radioactive contaminant uptake in on-site vegetation were identified. This may have resulted in probable contaminant migration and become introduced into the local ecological food chain. As a result, the SRS will initiate a time critical removal action in accordance with Section 300.415 of the NCP and FFA Section XIV to remove, treat (if required), and dispose of contaminated vegetation from the C-Reactor Seepage Basins. Erosion in the affected areas will be managed by an approved erosion control plan. further remediation of this unit will be conducted in accordance with the FFA

  8. Removal Site Evaluation Report to the C-Reactor Seepage Basins (904-066, -067 and -068G)

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, E.R. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-07-01

    Removal Site Evaluation Reports are prepared in accordance with Section 300.410 of the National Contingency Plan (NCP) and Section X of the Federal Facility Agreement (FFA). The C-Reactor Seepage Basins (904-066G,-067G,-068G) are listed in Appendix C, Resource Conservation and Recovery Act (RCRA)/Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Units List, of the FFA. The purpose of this investigation is to report information concerning conditions at this unit sufficient to assess the threat (if any) posed to human health and the environment and to determine the need for additional CERCLA action. The scope of the investigation included a review of past survey and investigation data, the files, and a visit to the unit.Through this investigation unacceptable conditions of radioactive contaminant uptake in on-site vegetation were identified. This may have resulted in probable contaminant migration and become introduced into the local ecological food chain. As a result, the SRS will initiate a time critical removal action in accordance with Section 300.415 of the NCP and FFA Section XIV to remove, treat (if required), and dispose of contaminated vegetation from the C-Reactor Seepage Basins. Erosion in the affected areas will be managed by an approved erosion control plan. further remediation of this unit will be conducted in accordance with the FFA.

  9. Enhanced treatment of wastewater from the vitamin C biosynthesis industry using a UASB reactor supplemented with zero-valent iron.

    Science.gov (United States)

    Shi, Rongjiu; Xu, Hui; Zhang, Ying

    2011-12-01

    The effects of zero-valent iron (Fe0) on the performance of a mesophilic upflow anaerobic sludge blanket (UASB) reactor treating high-strength wastewater from the vitamin C biosynthesis industry (VCW) was investigated during a 200-day period. The results showed that the chemical oxygen demand (COD) removal efficiency, CH4 content in biogas, specific methanogenic activity of sludge, and phosphate removal efficiency were significantly improved up to 81.8-96.1%, 76.5-79.6%, 1.71-2.87 g CH4-COD g(-1) VSS d(-1) and 68.5-85.2%, respectively, at elevated organic loading rates (OLRs) in the Fe0-amended reactor (RFe). In contrast, the corresponding values of 65.3-83.4%, 69.1-70.8%, 1.12-1.95 g CH4-COD g(-1) VSS d(-1) and 1.4-1.6%, respectively, were recorded in the control (R0). Elevated ferrous concentration of nearly 400 mg L(-1) in sludge was detected in RFe, whereas in the effluent of both reactors it was low (< 1.0 mg L(-1)). Batch tests further showed that Fe0 significantly enhanced the biodegradability of the VCW as shown by an increase in BOD/COD ratio from 0.41 to 0.65, and could serve as the electron donor for methanogenesis by anaerobic sludge, which were responsible for the differences between RFe and R0. The results suggest this integrated Fe0-microbial system is promising in facilitating the anaerobic digestion of VCW in UASB reactors. PMID:22439574

  10. Effect of temperature increase from 55 to 65 degrees C on performance and microbial population dynamics of an anaerobic reactor treating cattle manure

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær; Ibrahim, Ashraf; Mladenovska, Zuzana

    2001-01-01

    The effect of a temperature increase from 55 to 65 degreesC on process performance and microbial population dynamics were investigated in thermophilic, lab-scale, continuously stirred tank reactors. The reactors had a working volume of 3 l and were fed with cattle manure at an organic loading rate...... of 3 g VS/l reactor volume/d. The hydraulic retention time in the reactors was 15 days. A stable reactor performance was obtained for periods of three retention times both at 55 degreesC and 65 degreesC. At 65 degreesC methane yield stabilized at approximately 165 ml/g VS/d compared to 200 ml/g VS....../d at 55 degreesC. Simultaneously, Ibe level of total volatile fatty acids, VFA, increased from being below 0.3g/l to 1.8-2.4g acetate/l. The specific methanogenic activities (SMA) of biomass from the reactors were measured with acetate, propionate, butyrate, hydrogen, formate and glucose. At 65 degrees...

  11. Nanocrystalline SiC and Ti3SiC2 Alloys for Reactor Materials: Thermal and Mechanical Properties

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H.; Alvine, Kyle J.; Roosendaal, Timothy J.; Shin, Yongsoon; Nguyen, Ba Nghiep; Borlaug, Brennan A.; Jiang, Weilin

    2014-04-01

    SiC-polymers (pure polycarbosilane and polycarbosilane filled with SiC-particles) are being combined with Si and TiC powders to create a new class of polymer-derived ceramics for consideration as advanced nuclear materials in a variety of applications. Compared to pure SiC these materials have increased fracture toughness with only slightly reduced thermal conductivity. Future work with carbon nanotube (CNT) mats will be introduced with the potential to increase the thermal conductivity and the fracture toughness. At present, this report documents the fabrication of a new class of monolithic polymer derived ceramics, SiC + SiC/Ti3SiC2 dual phase materials. The fracture toughness of the dual phase material was measured to be significantly greater than Hexoloy SiC using indentation fracture toughness testing. However, thermal conductivity of the dual phase material was reduced compared to Hexoloy SiC, but was still appreciable, with conductivities in the range of 40 to 60 W/(m K). This report includes synthesis details, optical and scanning electron microscopy images, compositional data, fracture toughness, and thermal conductivity data.

  12. Analysis of Reference Design for Nuclear-Assisted Hydrogen Production at 750 C Reactor Outlet Temperature

    International Nuclear Information System (INIS)

    The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using a high-temperature gas-cooled reactor (HTGR) to provide the process heat and electricity to drive the electrolysis process. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This report describes the resulting new INL reference design coupled to two alternative HTGR power conversion systems, a Steam Rankine Cycle and a Combined Cycle (a Helium Brayton Cycle with a Steam Rankine Bottoming Cycle). Results of system analyses performed to optimize the design and to determine required plant performance and operating conditions when coupled to the two different power cycles are also presented. A 600 MWt high temperature gas reactor coupled with a Rankine steam power cycle at a thermal efficiency of 44.4% can produce 1.85 kg/s of hydrogen and 14.6 kg/s of oxygen. The same capacity reactor coupled with a combined cycle at a thermal efficiency of 42.5% can produce 1.78 kg/s of hydrogen and 14.0 kg/s of oxygen.

  13. Resolution of the Task A-11 reactor-vessel materials-toughness safety issue. Appendices C-K

    International Nuclear Information System (INIS)

    The central problem in the unresolved safety issue A-11, Reactor Vessel Materials Toughness, was to provide guidance in performing analyses required by 10 CFR Part 50, Appendix G, Section V.C. for reactor pressure vessels (RPVs) which fail to meet the toughness requirement during service life as a result of neutron radiation embrittlement. Although the methods of linear-elastic fracture mechanics (LEFM) were adequate for low-temperature RPV problems, they were inapplicable under operating conditions because vessel steels, even those which exhibit less than 50 ft-lb of C/sub v/ energy, were relatively tough at temperatures where the impact energy reached its upper shelf values. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which had been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the problem of RPV fracture with assumed beltline region flaws. The first paper of this report is a summary of the problem, the solutions, and the results of verification analyses. The details are provided in a series of appendices in Volumes I and II

  14. Multi-wafer 3C-SiC thin films grown on Si (100) in a vertical HWLPCVD reactor*

    Institute of Scientific and Technical Information of China (English)

    Yan Guoguo; Sun Guosheng; Wu Hailei; Wang Lei; Zhao Wanshun; Liu Xingfang; Zeng Yiping; Wen Jialiang

    2011-01-01

    We report the latest results of the 3C-SiC layer growth on Si (100) substrates by employing a novel home-made horizontal hot wall low pressure chemical vapour deposition (HWLPCVD) system with a rotating susceptor that was designed to support up to three 50 mm-diameter wafers. 3C-SiC film properties of the intrawafer and the wafer-to-wafer, including crystalline morphologies and electronics, are characterized systematically.Intra-wafer layer thickness and sheet resistance uniformity (σ/mean) of~3.40% and ~5.37% have been achieved in the 3 × 50 mm configuration. Within a run, the deviations of wafer-to-wafer thickness and sheet resistance are less than 4% and 4.24%, respectively.

  15. Study on the first wall TiC coated materials for fusion reactor

    International Nuclear Information System (INIS)

    The chemical vapor deposition (CVD) process of TiC coating, electron beam thermal shock and thermal fatigue testing of TiC coated materials are described. The dense and fine coating is deposited at 1100 degree, CH4 flux of 0.36 L/min and H2 flux of 1.16 L/min, and the deposition rate reaches 0.7 μm/min. The correlation between coating thickness and process parameters is given. Pulsed by electron beams with high power density up to 226 MW/m2 for 0.6 s, the TiC layers of TiC/graphite, TiC/molybdenum and TiC/316L SS spall from substrates, and 316L SS is molten. A lot of TiC layer spall from 316L SS after 2 hear cycles between 900 degree C and -246 degree C, net-cracks are formed on the surface of TiC/graphite during the fatigue testing, but no exfoliation of TiC layer is observed up to the maximum heat cycles 200. Neither cracks nor exfoliation of TiC layer on molybdenum are found after 200 heat cycles

  16. Status quo of the storage of C-14 containing waste from German high-temperature and research reactors and the strategy for their future storage

    International Nuclear Information System (INIS)

    The German research reactors and the high-temperature reactors (AVR, THTR 300) used graphite as reflector and/or moderator materials. Due to the activation of these materials there exists a significant amount of C-14, H-3 and Co-60.The project included the assessment of the respective amount of materials from all facilities, the legal framework for interim storage, transport and final disposal.

  17. Effect of high salinity on the fate of methanol during the start-up of thermophilic (55°C) sulfate reducing reactors

    NARCIS (Netherlands)

    Vallero, M.V.G.; Hulshoff Pol, L.W.; Lens, P.N.L.; Lettinga, G.

    2002-01-01

    Two 6.5 L lab-scale upflow anaerobic sludge bed (UASB) reactors were operated at 55°C fed with methanol as the sole electron and carbon source and in excess of sulfate (COD/SO42- of 0.5) in order to investigate the effect of high wastewater salinity on the start-up period. The first reactor (UASB I)

  18. Residual stress in the first wall coating materials of TiC and TiN for fusion reactor

    International Nuclear Information System (INIS)

    Residual stresses measurement in the first wall coating of a fusion reactor of TiC and TiN films by X-ray diffraction 'sin2ψ methods' were described. The authors have studied on the effect of conditions of specimen preparation (such as coating method, substrate materials, film thickness and deposition temperature) on the residual stress of TiC and TiN films coated onto Mo, 316LSS and Pocographite by chemical vapor deposition (CVD) and physical vapor deposition (PVD) method. All films prepared in this study were found to have a compressive stresses and the CVD method gave lower residual stress than PVD method. TiC film coated on Mo substrate at 1100 degree C by CVD method showed that residual stress as the film thickness was raised from 14 μm to 60 μm, on the other hand, residual stress by PVD method exhibited a high compressive stresses, this kind of stress was principally the intrinsic stress, and a marked decrease in the residual with raising the deposition temperature (200 degree C∼650 degree C) was demonstrated. Origins of the residual stress were discussed by correlation with differences between thermal expansion coefficients, and also with fabrication methods

  19. PRAISE-C, Double-Ended Guillotine Break (DEGB) Breaks from Weld Cracks in Light-Water Reactor Piping System

    International Nuclear Information System (INIS)

    1 - Description of program or function: PRAISE-C is a probabilistic fracture mechanics code used to estimate the probability of double-ended guillotine break (DEGB) in light water reactor piping due to the growth of cracks at welded joints. Pipe failures are considered to occur as the result of crack-like defects either introduced during fabrication, or that initiate after plant operation has begun, and that escape detection during inspections. PRAISE was developed to estimate the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. An initial hydrostatic proof test, pre-service non-destructive inspection, and periodic in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients, such as system heatup and cooldown, either as constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. Earthquakes of varying intensity and arbitrary occurrence times can be modeled. PRAISE-C extends the capabilities of PRAISE-B to include a tearing instability failure criterion for carbon steels (supplementing the original net section stress criterion used for austenitic materials), and an advanced probabilistic model of stress corrosion cracking in stainless steels (Type 304, Type 316NG 'nuclear grade') used for BWR reactor coolant piping. The stress corrosion model is semi-empirical in nature, and is based on experimental and field data. The model considers crack initiation, including the number, time, and location of initiated cracks, in addition to the effect of stress corrosion on crack growth rates. Various phenomena are considered, including environment (i.e., coolant temperature, dissolved oxygen content, level of impurities), applied loads, residual stresses, material type, and degree of sensitization. By allowing cracks to initiate after reactor operation has begun, the simulation

  20. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  1. Synthesis of SiC from rice husk in a plasma reactor

    Indian Academy of Sciences (India)

    S K Singh; B C Mohanty; S Basu

    2002-11-01

    A new route for production of SiC from rice husk is reported by employing thermal plasma technique. The formation of -SiC is observed in a short time of 5 min. The samples are characterized by XRD and SEM.

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

    International Nuclear Information System (INIS)

    A continuous stirred tank reactor (CSTR) (750 cm3 working volume) was operated with pig slurry under hyper-thermophilic (70 oC) temperature for hydrogen production. The hydraulic retention time (HRT) was 24 h and the organic loading rate was 24.9 g d-1 of volatile solid (VS). The inoculum used in the hyper-thermophilic reactor was sludge obtained from a mesophilic methanogenic reactor. The continuous feeding with active biomass (inoculum) from the mesophilic methanogenic reactor was necessary in order to achieve hydrogen production. The hyper-thermophilic reactor started to produce hydrogen after a short adapted period of 4 days. During the steady state period the mean hydrogen yield was 3.65 cm3 g-1 of volatile solid added. The high operation temperature of the reactor enhanced the hydrolytic activity in pig slurry and increased the volatile fatty acids (VFA) production. The short HRT (24 h) and the hyper-thermophilic temperature applied in the reactor were enough to prevent methanogenesis. No pre-treatment methods or other control methods for preventing methanogenesis were necessary. Hyper-thermophilic hydrogen production was demonstrated for the first time in a CSTR system, fed with pig slurry, using mixed culture. The results indicate that this system is a promising one for biohydrogen production from pig slurry.

  3. Chemical compatibility of SiC composite structures with fusion reactor helium coolant at high temperatures

    International Nuclear Information System (INIS)

    The thermodynamic stability of SiC/SiC composite structures proposed for fusion applications is presented in this paper. Minimization of the free energy for reacting species in the temperature range 773-1273 K is achieved by utilizing the NASA-Lewis Chemical Equilibrium Thermodynamics Code (CET). The chemical stability of the matrix (SiC), as well as several fiber coatings (BN and graphite) are studied. Helium coolant is assumed to contain O2 and water moisture impurities in the range 100-1000 ppm. The work is applied to recent Magnetic and Inertial Confinement Conceptual designs. The present study indicates that the upper useful temperature limit for SiC/SiC composites, from the standpoint of high-temperature corrosion, will be in the neighborhood of 1273 K. Up to this temperature, corrosion of SiC is shown to be negligible. The main mechanism of weight loss will be by evaporation to the plasma side. The presence of a protective SiO2 condensed phase is discussed, and is shown to result in further reduction of high-temperature corrosion. The thermodynamic stability of C and BN is shown to be very poor under typical fusion reaction conditions. Further development of chemically stable interface materials is required. (orig.)

  4. Development of Double Rotation C-Scanning System and Program for Under-Sodium Viewing of Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    A double rotation C-scanning system and a software program Under-Sodium MultiVIEW have been developed for the under-sodium viewing of a reactor core and in-vessel structures of a sodium-cooled fast reactor KALIMER-600. Double rotation C-scanning system has been designed and manufactured by the reproduction of double rotation plug of a reactor head in KALIMER-600. Hardware system which consists of a double rotating scanner, ultrasonic waveguide sensors, a high power ultrasonic pulser-receiver, a scanner driving module and a multi channel A/D board have been constructed. The functions of scanner control, image mapping and signal processing of Under-Sodium MultiVIEW program have been implemented by using a LabVIEW graphical programming language. The performance of Under-Sodium MultiVIEW program was verified by a double rotation C-scanning test in water

  5. Application of macro-cellular SiC reactor to diesel engine-like injection and combustion conditions

    OpenAIRE

    Cypris, Jochen; Weclas, Miroslaw; Greil, Peter; Schlier, Lorenz M.; Travitzky, Nahum; Zhang, W

    2012-01-01

    One of novel combustion technologies for low emissions and highly efficient internal combustion engines is combustion in porous reactors (PM). The heat release process inside combustion reactor is homogeneous and flameless resulting in a nearly zero emissions level. Such combustion process, however is non-stationary, is performed under high pressure with requirement of mixture formation directly inside the combustion reactor (high pressure fuel injection). Reactor heat capacity resulting in l...

  6. Reactivity and neutron flux measurements in IPEN/MB-01 reactor with B4C burnable poison

    International Nuclear Information System (INIS)

    Burnable poison rods, made of B4C- Al2 O3 pellets with 5.01 mg/cm310 B concentration, have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. Several core parameters which are affected by the burnable poisons rods have been measured. The principal results, for the situation in which the burnable poison rods are located near the absorber rods of a control rod, are they cause a 29% rod worth shadowing, a reduction of 39% in the local void coefficient of reactivity, a reduction of 4.8% in the isothermal temperature coefficient of reactivity, and a reduction of 9% in the thermal neutron flux in the region where the burnable poison rods are located. These experimental results will be used for the validation of burnable poison calculation methods in the CTMSP. (author)

  7. Exchange of pipes of the reactor water clean up system in nuclear power plant KRB II B and C

    International Nuclear Information System (INIS)

    The original pipes of the reactor water clean up system in KRB II were made of the Ti-stabilised austenitic steel 1.4541 (X6CrNiTi1810). After 10 years of operation, non-destructive examinations detected cracks in the welds. The metallograohic investigations revealed the well-known intergranular crack paths. The flawed welded sections therefore were replaced, and it was decided to completely replace the outlet pipes of the two units, between RPV nozzle (safe end) and the regenerative heat exchangers. This was done in unit B in the inspection period of 1996, and in unit C in 1997. The objective of the complete replacements was to basically enhance system safety in regard to cracking susceptibility of the material, loads, fatigue effects, and in-service inspection. The main aspects and effects of the replacements are discussed and explained. (orig./CB)

  8. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  9. A method of measuring αc on fast neutron reactor

    International Nuclear Information System (INIS)

    The power rise of burst leading edge is measured at super prompt criticality and prompt neutron breeding. constant (α) is gained through fitting the curve of power. We calibrate the reactivity of rod at super delay criticality and deduce the pre-load reactivity from the reactivity of rod. The prompt neutron decay constant at delay criticality (αc) and directional difference on CFBR-Ⅱ are got from deduction of experimental data at super prompt criticality. The measured αc accords with the result of Rosii-α. (authors)

  10. Temperature increases from 55 to 75 C in a two-phase biogas reactor result in fundamental alterations within the bacterial and archaeal community structure

    Energy Technology Data Exchange (ETDEWEB)

    Rademacher, Antje [Leibniz-Institut fuer Agrartechnik Potsdam-Bornim e.V. (ATB), Potsdam (Germany). Abt. Bioverfahrenstechnik; Technische Univ. Berlin (Germany). Inst. fuer Technischen Umweltschutz; Nolte, Christine; Schoenberg, Mandy; Klocke, Michael [Leibniz-Institut fuer Agrartechnik Potsdam-Bornim e.V. (ATB), Potsdam (Germany). Abt. Bioverfahrenstechnik

    2012-10-15

    Agricultural biogas plants were operated in most cases below their optimal performance. An increase in the fermentation temperature and a spatial separation of hydrolysis/acetogenesis and methanogenesis are known strategies in improving and stabilizing biogas production. In this study, the dynamic variability of the bacterial and archaeal community was monitored within a two-phase leach bed biogas reactor supplied with rye silage and straw during a stepwise temperature increase from 55 to 75 C within the leach bed reactor (LBR), using TRFLP analyses. To identify the terminal restriction fragments that were obtained, bacterial and archaeal 16S rRNA gene libraries were constructed. Above 65 C, the bacterial community structure changed from being Clostridiales-dominated toward being dominated by members of the Bacteroidales, Clostridiales, and Thermotogales orders. Simultaneously, several changes occurred, including a decrease in the total cell count, degradation rate, and biogas yield along with alterations in the intermediate production. A bioaugmentation with compost at 70 C led to slight improvements in the reactor performance; these did not persist at 75 C. However, the archaeal community within the downstream anaerobic filter reactor (AF), operated constantly at 55 C, altered by the temperature increase in the LBR. At an LBR temperature of 55 C, members of the Methanobacteriales order were prevalent in the AF, whereas at higher LBR temperatures Methanosarcinales prevailed. Altogether, the best performance of this two-phase reactor was achieved at an LBR temperature of below 65 C, which indicates that this temperature range has a favorable effect on the microbial community responsible for the production of biogas. (orig.)

  11. Photocatalytic membrane reactor for the removal of C.I. Disperse Red 73

    OpenAIRE

    Valentina Buscio; Stephan Brosillon; Julie Mendret; Martí Crespi; Carmen Gutiérrez-Bouzán

    2015-01-01

    After the dyeing process, part of the dyes used to color textile materials are not fixed into the substrate and are discharged into wastewater as residual dyes. In this study, a heterogeneous photocatalytic process combined with microfiltration has been investigated for the removal of C.I. Disperse Red 73 from synthetic textile effluents. The titanium dioxide (TiO2) Aeroxide P25 was selected as photocatalyst. The photocatalytic treatment achieved between 60% and 90% of dye degradation and up ...

  12. Development of Double Rotation C-Scan Program for the Visualization of a Reactor Core and In-vessel Structures in SFR

    International Nuclear Information System (INIS)

    The in-service inspection (ISI) of in-vessel structures of a sodium-cooled fast reactor (SFR) should be applied for the periodic safety assurance of SFR according to the ASME code rules and guidelines. As the liquid sodium of a sodium-cooled fast reactor (SFR) is opaque to light, a conventional visual inspection is unavailable for the inspection of the reactor core and in-vessel structures under a sodium level. An ultrasonic wave should be applied for an under-sodium viewing (USV) of the in-vessel structures. Under-sodium ultrasonic sensors have been widely developed for an ISI of the reactor core and in-vessel structures of SFR. In previous research works, the novel plate waveguide sensor has been developed for the application of remote under-sodium viewing and also the feasibility of the waveguide sensor technique has been successfully demonstrated for an ultrasonic visual inspection of a reactor core and in-vessel structures. KALIMER-600 has a double rotating plug in reactor head for the refueling of reactor fuel assemblies. The waveguide sensors will be permanently installed in the double rotating plug for the mapping of reactor core and in-service inspection of in-vessel structures. In this study, a double rotating scanner system and a C-scan image mapping program Under-Sodium MultiVIEW (US-MultiVIEW) have been developed to apply this waveguide sensor technology to an under-sodium inspection of in-vessel structures of SFR by using a LabVIEW (Laboratory Virtual Instrumentation Engineer.s Workbench) graphical programming language

  13. Use of SiCf/SiC ceramic composites as structure material of a fusion reactor toroid internal components

    International Nuclear Information System (INIS)

    The use of low neutron-induced activation structural materials seems necessary in order to improve safety in future fusion power reactors. Among them, SiCf/SiC composites appear as a very promising solution because of their low activation characteristics coupled with excellent mechanical properties at high temperatures. With the main objective of evaluating the limit of present-day composites, a tritium breeding blanket using SiCf/SiC as structural material (the TAURO blanket) has been developed in the last years by the Commissariat a l'Energie Atomique (CEA). The purpose of this thesis was to modify the available design tools (computer codes, design criteria), normally used for the analyses of metallic structures, in order to better take into account the mechanical behaviour of SiCf/SiC. Alter a preliminary improvement of the calculation methods, two main topics of study could be identified: the modelling of the mechanical behaviour of the composite and the assessment of appropriate design criteria. The different behavioural models available in literature were analysed in order to find the one that was the best suited to the specific problems met in the field of fusion power. The selected model was then implemented in the finite elements code CASTEM 2000 used within the CEA for the thermo-mechanical analyses of the TAURO blanket. For the design of the blanket, we proposed a new resistance criterion whose main advantage, with respect to the other examined, lies in the easiness of identification. The suggested solutions were then applied in the design studies of the TAURO blanket. We then could show that the use of appropriate calculation methodologies is necessary in order to achieve a correct design of the blanket and a more realistic estimate of the limits of present day composites. The obtained results can also be extended to all nuclear components making use of SiCf/SiC structures. (author)

  14. High density LHRF experiments in Alcator C-Mod and implications for reactor scale devices

    Science.gov (United States)

    Baek, S. G.; Parker, R. R.; Bonoli, P. T.; Shiraiwa, S.; Wallace, G. M.; LaBombard, B.; Faust, I. C.; Porkolab, M.; Whyte, D. G.

    2015-04-01

    Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the ‘density limit’, and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent

  15. Microbial succession within an anaerobic sequencing batch biofilm reactor (ASBBR treating cane vinasse at 55ºC

    Directory of Open Access Journals (Sweden)

    Maria Magdalena Ferreira Ribas

    2009-08-01

    Full Text Available The aim of this work was to investigate the anaerobic biomass formation capable of treating vinasse from the production of sugar cane alcohol, which was evolved within an anaerobic sequencing batch biofilm reactor (ASBBR as immobilized biomass on cubes of polyurethane foam at the temperature of 55ºC. The reactor was inoculated with mesophilic granular sludge originally treating poultry slaughterhouse wastewater. The evolution of the biofilm in the polyurethane foam matrices was assessed during seven experimental phases which were thus characterized by the changes in the organic matter concentrations as COD (1.0 to 20.0 g/L. Biomass characterization proceeded with the examination of sludge samples under optical and scanning electron microscopy. The reactor showed high microbial morphological diversity along the trial. The predominance of Methanosaeta-like cells was observed up to the organic load of 2.5 gCOD/L.d. On the other hand, Methanosarcinalike microorganisms were the predominant archaeal population within the foam matrices at high organic loading ratios above 3.3 gCOD/L.d. This was suggested to be associated to a higher specific rate of acetate consumption by the later organisms.Este trabalho investigou a formação de um biofilme anaeróbio capaz de tratar vinhaça da produção de álcool de cana-de-açúcar, que evoluiu dentro de um reator operado em bateladas seqüenciais com biofilme (ASBBR tendo a biomassa imobilizada em cubos de espuma de poliuretano na temperatura de 55ºC. O reator foi inoculado com lodo granular mesofílico tratando água residuária de abatedouro de aves. A evolução do biofilme nas matrizes de espuma de poliuretano foi observada durante sete fases experimentais que foram caracterizadas por mudanças nas concentrações de matéria orgânica como DQO (1,0 a 20,0 g/L. A caracterização da biomassa foi feita por exames de amostras do lodo em microscopia ótica e eletrônica de varredura. O reator apresentou

  16. Photocatalytic Membrane Reactor for the Removal of C.I. Disperse Red 73

    Directory of Open Access Journals (Sweden)

    Valentina Buscio

    2015-06-01

    Full Text Available After the dyeing process, part of the dyes used to color textile materials are not fixed into the substrate and are discharged into wastewater as residual dyes. In this study, a heterogeneous photocatalytic process combined with microfiltration has been investigated for the removal of C.I. Disperse Red 73 from synthetic textile effluents. The titanium dioxide (TiO2 Aeroxide P25 was selected as photocatalyst. The photocatalytic treatment achieved between 60% and 90% of dye degradation and up to 98% chemical oxygen demand (COD removal. The influence of different parameters on photocatalytic degradation was studied: pH, initial photocatalyst loading, and dye concentration. The best conditions for dye degradation were pH 4, an initial dye concentration of 50 mg·L−1, and a TiO2 loading of 2 g·L−1. The photocatalytic membrane treatment provided a high quality permeate, which can be reused.

  17. Verification of the shift Monte Carlo code with the C5G7 reactor benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Sly, N. C.; Mervin, B. T. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States); Mosher, S. W.; Evans, T. M.; Wagner, J. C. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)

  18. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  19. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content

  20. Mechanical properties and microstructure of three Russian ferritic/martensitic steels irradiated in BN-350 reactor to 50 dpa at 490 oC

    International Nuclear Information System (INIS)

    Ferritic/martensitic (F/M) steels are being considered for application in fusion reactors, intense neutron sources, and accelerator-driven systems. While EP-450 is traditionally used with sodium coolants in Russia, EP-823 and EI-852 steels with higher silicon levels have been developed for reactor facilities using lead-bismuth coolant. To determine the influence of silicon additions on short-term mechanical properties and microstructure, ring specimens cut from cladding tubes of these three steels were irradiated in sodium at 490 oC in the BN-350 reactor to 50 dpa. Post-irradiation tensile testing and microstructural examination show that EI-852 steel (1.9 wt% Si) undergoes severe irradiation embrittlement. Microstructural investigation showed that the formation of near-continuous χ-phase precipitates on grain boundaries is the main cause of the embrittlement

  1. Screening for potential fermentative hydrogen production from black water and kitchen waste in on-site UASB reactor at 20 degrees C.

    Science.gov (United States)

    Luostarinen, S; Pakarinen, O; Rintala, J

    2008-06-01

    The potential of black water and a mixture of black water and kitchen waste as substrates for on-site dark fermentative hydrogen production was screened in upflow anaerobic sludge blanket reactors at 20 degrees C. Three different inocula were used with and without heat treatment. With glucose, the highest specific hydrogenogenic activity was 69 ml H2 g volatile solids(-1) d(-1) in batch assays and the highest hydrogen yield 0.44 mol H2 mol glucose(-1) in upflow anaerobic sludge bed reactor. The mixture of black water and kitchen waste degraded readily into volatile fatty acids in the reactors, thus showing potential for hydrogen production. In the conditions applied, however, the highest end product was propionate and no hydrogen was produced. Black water alone apparently contained too little readily soluble carbohydrates for hydrogen producing bacteria, and little VFA and no hydrogen was produced. PMID:18702295

  2. One- and two-stage upflow anaerobic sludge-bed reactor pretreatment of winery wastewater at 4-10 degreesC.

    Science.gov (United States)

    Kalyuzhnyi, S V; Gladchenko, M A; Sklyar, V I; Kizimenko, Y S; Shcherbakov, S S

    2001-02-01

    The operating performance of a single and two (in series) laboratory upflow anaerobic sludge-bed (UASB) reactors (2.7-L working volume, recycle ratio varied from 1:1 to 1:18) treating diluted wine vinasse was investigated under psychrophilic conditions (4-10 degreesC). For a single UASB reactor seeded with granular sludge, the average organic loading rates (OLRs) applied were 4.7, 3.7, and 1.7 g of chemical oxygen demand (COD)/(L.d) (hydraulic retention times [HRTs] were about 1 d) at 9-11, 6 to 7, and 4 to 5 degreesC, respectively. The average total COD removal for preacidified vinasse wastewater was about 60% for all the temperature regimes tested. For two UASB reactors in series, the average total COD removal for treatment of non-preacidified wastewater exceeded 70% (the average OLRs for a whole system were 2.2, 1.8, and 1.3 g of COD/[L.d] under HRTs of 2 d at 10, 7, and 4 degreesC, respectively). In situ determinations of kinetic sludge characteristics (apparent Vm and Km) revealed the existence of substantial mass transfer limitations for the soluble substrates inside the reactor sludge bed. Therefore, application of higher recycle ratios is essential for enhancement of UASB pretreatment under psychrophilic conditions. The produced anaerobic effluents were shown to be efficiently posttreated aerobically: final effluent COD concentrations were about 0.1 g/L. Successful operation of the UASB reactors at quite low temperatures (4-10 degreesC) opens some perspectives for application of high-rate anaerobic pretreatment at ambient temperatures. PMID:11297387

  3. Interaction of Impurity (Li, Be, B and C)and Hydrogen Isotope Pellet Injection with Reactor-relevant Plasmas

    Institute of Scientific and Technical Information of China (English)

    Deng Baiquan(邓柏权); J.P.Allain; Peng Lilin(彭利林); Wang Xiaoyu(王晓宇); Chen Zhi(陈志); Yan Jiancheng(严建成)

    2005-01-01

    Based on the two-dimensional kinetic ablation theory of the hydrogen pellet ablation developed by Kuteev [B.V. Kuteev, Nuclear Fusion, 35 (1995) 431], an algorithm of erosion speed and ablation rate calculations for Li, Be, and B impurity pellets in reactor-relevant plasma has been derived. Results show compatibilities of lithium pellet injection used in α-particle diagnostics are positive in comparison with other solid impurity pellets (e.g. Be, B and C). Using the 2-D Kuteev lentil model, including kinetic effects, we find that currently existing pellet injection techniques will not meet core-fueling requirements for ITER-FEAT. A pressure as high as 254 MPa must be applied to a pellet accelerator with a 200 cm-long single-stage pneumatic gun, in order to accelerate a pellet with a radius rp0 =0.5 cm to a velocity of Vp0, 24×105 cm/s penetrating 100 cm into the ITER plasma core. Comparisons of pellet velocity- and radius-dependent penetration depth between the Neutral Gas Shielding and the Kuteev's models are made. However, we find that the isotopic effects can lead to a 33% lower pellet speed for solid DT, compared to an identical H2 pellet penetrating the same length in ITER-FEAT plasma, and our calculations show that HFS injection will much improve core fueling efficiency.

  4. Study the effects of different reflector types on the neutronic parameters of the 10 MW MTR reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    Highlights: • A 3-D neutronic model for the 10 MW MTR has been conducted using the MCNP4C code. • Studying the effect of different reflectors on the neutronics parameters of the reactor. • Beryllium reflector was found to be the most efficient reflector among the studied reflectors. • The graphite reflector gave the highest maximum thermal neutron flux in the water trap. - Abstract: A 3-D neutronic model for the 10 MW MTR research reactor has been conducted for the HEU (93%), MEU (45%) and LEU (20%) fuels using the MCNP4C code. This model has been used to study the effect of different types of reflector materials on the reactor multiplication factor and neutron flux distribution in the reactor. It was found that the beryllium reflector was the most efficient reflector among the studied reflector groups (beryllium, heavy water, graphite and light water) since it gave the highest reactor multiplication factor, 1.21441. It followed by heavy water, graphite and light water with the following reactor multiplication factors: 1.19458, 1.19287 and 1.16867 respectively. The graphite reflector gave the highest maximum thermal neutron flux in the water trap, 2.576E14 n cm−2 s−1. It followed by heavy water, light water, and beryllium with the following results: 2.533E14, 2.526E14 and 2.525E14 n cm−2 s−1 respectively. Considerable gains in reactivity were not appreciably influenced by changing the fuel enrichment

  5. Minimizing MTR reactor uranium load with the use of MOX fuel by employing ORIGEN-S and MCNP4C codes

    International Nuclear Information System (INIS)

    Highlights: • Recycling of the ETRR-2 by MOX fuel elements. • Calculation of the neutronic parameters of the ETRR-2 after recycling by MOX fuel elements. • MCNP5-beta code is coupled with ORIGEN-S by a set of interface programs. - Abstract: A computational study was performed for MTR reactors using ORIGEN-S and MCNP4C codes to replace some of the fuel elements (FEs) with MOX FEs. The results show that the replacement of the MTR-22 MW power research reactor Fes with MOX FEs leads to the reduction in the enrichment with 235U and the amount of loaded 235U in the core up to more than 20%. The amount of loaded uranium 235U FEs decreased considerably by increasing the number of MOX FEs. Re-evaluated neutronic parameters of the reactor showed that the replacement of the FEs by MOX FEs does not affect negatively the safe operational conditions of the reactor with practically no harmful effect on the safety of the reactor

  6. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  7. Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [ORNL; Fugate, David L [ORNL; Cetiner, Sacit M [ORNL; Qualls, A L [ORNL

    2015-05-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactor innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  8. A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)

    Science.gov (United States)

    A dynamic kinetic model for the advanced oxidation process (AOP) using hydrogen peroxide and ultraviolet irradiation (H2O2/UV) in a completely mixed batch reactor (CMBR) is developed. The model includes the known elementary chemical and photochemical reac...

  9. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  10. Process heat reactors

    International Nuclear Information System (INIS)

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  11. Effect of dissolved hydrogen on the corrosion behavior of chemically vapor deposited SiC in a simulated pressurized water reactor environment

    International Nuclear Information System (INIS)

    Highlights: • Corrosion behavior of CVD SiC was studied under PWR-simulating water conditions. • Dissolved hydrogen significantly reduced the corrosion rate of SiC. • Dissolved hydrogen was likely to retard the formation of surface oxide layer. - Abstract: The corrosion behavior of chemically vapor deposited SiC was investigated in relation to its application as fuel cladding in a pressurized water reactor (PWR). Corrosion tests were conducted in deoxygenated, pressurized water at 360 °C with control of dissolved hydrogen. The dissolved hydrogen dramatically reduced the corrosion rate of SiC compared with that in water without the dissolved hydrogen. It was revealed that the dissolution of the surface oxide of SiC primarily contributed to the weight loss at the initial stage of corrosion. Further weight loss was minimal for 210 d because dissolved hydrogen effectively retarded the formation of SiO2

  12. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    John Collins

    2009-08-01

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  13. Investigation of plant control strategies for the supercritical C02Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code

    International Nuclear Information System (INIS)

    The development of a control strategy for the supercritical CO2 (S-CO2) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO2 Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO2 Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO2 heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO2 cycle conditions adjust according to the S-CO2 cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate inlet sodium temperatures by about 10 C

  14. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    International Nuclear Information System (INIS)

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called U-batteryTM, which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  15. Determination of βeff using MCNP-4C2 and application to the CROCUS and PROTEUS reactors

    International Nuclear Information System (INIS)

    A new Monte Carlo method for the determination of βeff has been recently developed and tested using appropriate models of the experimental reactors CROCUS and PROTEUS. The current paper describes the applied methodology and highlights the resulting improvements compared to the simplest MCNP approach, i.e. the 'prompt method' technique. In addition, the flexibility advantages of the developed method are presented. Specifically, the possibility to obtain the effective delayed neutron fraction βeff per delayed neutron group, per fissioning nuclide and per reactor region is illustrated. Finally, the MCNP predictions of βeff are compared to the results of deterministic calculations. (authors)

  16. Analysis of gamma ray intensity on the S/C vent pipes area in the unit 2 reactor building of the Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The robot is equipped with cameras, a dosimeter, and 2 DOF (degree of freedom) manipulation arms. It loads a small vehicle equipped with a camera that can access and inspect narrow areas. TEPCO is using the four-legged walking robot to inspect the suppression chamber (S/C) area of the unit 2 reactor building basement in the Fukushima Daiichi Nuclear Power Plant. The robot carried out 6 missions for about four months, from 11 December, 2012 to 15 March, 2013, where it examined an evidence of a leakage of radioactivity contaminated water in the S/C area of unit 2 reactor building. When a camera's signal processing unit, which is consist of ASIC and FPGA devices manufactured by a CMOS fabrication process, is exposed to a higher dose rate gamma ray, the speckle distribution in the camera image increase more. From the inspection videos, released by TEPCO, of the underground 8 vent pipes in the unit 2 reactor building, we analyzed the speckle distribution from the high dose-rate gamma rays. Based on the distribution of the speckle, we attempted to characterize the vent pipe with much radioactivity contaminated materials among the eight vent pipes connected to the PCV. The numbers of speckles viewed in the image of a CCD (or CMOS) camera are related to an intensity of the gamma ray energy emitted by a nuclear fission reaction from radioactivity materials. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. Therefore, calculating the speckles counts, we can determine the vent pipe with relatively most radioactivity-contaminated materials among the other vent pipes. From the comparison of speckles counts calculated in the inspection image of the vent pipe with the speckles counts extracted by gamma ray irradiation experiment of the same small vehicle camera model loaded with the four-legged walking robot, we can qualitatively estimate the gamma ray dose-rate in the S/C vent pipe area of the

  17. Analysis of gamma ray intensity on the S/C vent pipes area in the unit 2 reactor building of the Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    The robot is equipped with cameras, a dosimeter, and 2 DOF (degree of freedom) manipulation arms. It loads a small vehicle equipped with a camera that can access and inspect narrow areas. TEPCO is using the four-legged walking robot to inspect the suppression chamber (S/C) area of the unit 2 reactor building basement in the Fukushima Daiichi Nuclear Power Plant. The robot carried out 6 missions for about four months, from 11 December, 2012 to 15 March, 2013, where it examined an evidence of a leakage of radioactivity contaminated water in the S/C area of unit 2 reactor building. When a camera's signal processing unit, which is consist of ASIC and FPGA devices manufactured by a CMOS fabrication process, is exposed to a higher dose rate gamma ray, the speckle distribution in the camera image increase more. From the inspection videos, released by TEPCO, of the underground 8 vent pipes in the unit 2 reactor building, we analyzed the speckle distribution from the high dose-rate gamma rays. Based on the distribution of the speckle, we attempted to characterize the vent pipe with much radioactivity contaminated materials among the eight vent pipes connected to the PCV. The numbers of speckles viewed in the image of a CCD (or CMOS) camera are related to an intensity of the gamma ray energy emitted by a nuclear fission reaction from radioactivity materials. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. Therefore, calculating the speckles counts, we can determine the vent pipe with relatively most radioactivity-contaminated materials among the other vent pipes. From the comparison of speckles counts calculated in the inspection image of the vent pipe with the speckles counts extracted by gamma ray irradiation experiment of the same small vehicle camera model loaded with the four-legged walking robot, we can qualitatively estimate the gamma ray dose-rate in the S/C vent pipe area of the

  18. System of quality assurance for the RP-10 research nuclear reactor ISO 9000 and SS No. 50-C/SG-Q

    International Nuclear Information System (INIS)

    The object of this report is the development of a Quality Assurance Manual expressing the system of quality assurance for the nuclear reactor; in this case, for the RP-10 Reactor. Two standards for the system of quality assurance are taken as reference standards: ISO 9001, as quality management standard and the Safety Series No. 50-C/SG-Q nuclear standard, as technical standard for nuclear facilities, which is mainly focused in the facility safety. A situational analysis of the installation was carried out, such as the diagnosis. Then, the improvement of the quality system of the installation and the actions to be performed are proposed. Finally, implementation of the new system of quality assurance, the documents to be considered and the quality assurance manual are shown

  19. Solar Hydrogen and C-Nanofilaments by Thermal Decomposition of CH{sub 4} Using 5kW Vortex-Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, D. [ETH Zuerich (Switzerland); Trommer, D. [ETHZ (Switzerland); Steinfeld, A. [ETHZ and PSI (Switzerland)

    2004-03-01

    The solar thermal decomposition of CH{sub 4} for co-producing hydrogen and carbon is experimentally investigated. The solar chemical reactor features a vortex flow of CH{sub 4} confined to a cavity-receiver and laden with carbon particles that serve simultaneously as radiant absorbers and nucleation sites. A 5 kW reactor prototype, tested at PSI's solar furnace and ETH's high-flux solar simulator with power flux intensities exceeding 3500 kW/m{sup 2}, yielded 67% chemical conversion of CH{sub 4} to H{sub 2} and C(gr) at 1600 K and 1 bar. Carbon formed was of nano-filamentary nature. The proposed solar hybrid chemical process conserves natural gas, reduces CO{sub 2} emissions, and provides a transition path to solar hydrogen. (author)

  20. Job/task analysis for I ampersand C [Instrumentation and Controls] instrument technicians at the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    To comply with Department of Energy Order 5480.XX (Draft), a job/task analysis was initiated by the Maintenance Management Department at Oak Ridge National Laboratory (ORNL). The analysis was applicable to instrument technicians working at the ORNL High Flux Isotope Reactor (HFIR). This document presents the procedures and results of that analysis. 2 refs., 2 figs

  1. Silicon Carbide Nanoparticles Produced by CO2 Laser Pyrolysis of SiH4/C2H2 Gas Mixtures in a Flow Reactor

    International Nuclear Information System (INIS)

    Pulsed CO2-laser-induced decomposition of different mixtures of SiH4 and C2H2 in a flow reactor has been employed to produce silicon carbide clusters and nanoparticles with varying content of carbon. The as-synthesized species were extracted from the reaction zone by a conical nozzle and expanded into the source chamber of a cluster beam apparatus where, after having traversed a differential chamber, they were analyzed with a time-of-flight mass spectrometer. Thin films of silicon carbide nanoclusters were produced by depositing the clusters at low energy on potassium bromide and sapphire windows mounted into the differential chamber. At the same time, Si and SiC nanoparticles were collected in a filter placed into the exhaust line of the flow reactor. Both beam and powder samples were characterized by FTIR spectroscopy. The close resemblance of the spectra suggests that the composition of the beam and powder particles obtained during the same run is nearly identical. XRD spectroscopy could only be employed for the investigation of the powders. It was found that CO2 laser pyrolysis is ideally suited to produce silicon carbide nanoparticles with a high degree of crystallinity. Nanopowders produced from the pyrolysis of a stoichiometric (2:1) mixture of SiH4/C2H2 were found to contain particles or domains of pure silicon. The characteristic silicon features in the FTIR and XRD spectra, however, disappeared when C2H2 was applied in excess

  2. Modelización del estado estacionario de las condiciones de flujo de un reactor TRIGA mediante el código de dinámica de fluidos computacional ANSYS CFX

    OpenAIRE

    Martínez Lianes, Mónica; Miró Herrero, Rafael; Verdú Martín, Gumersindo Jesús; Pereira, C; CHIVA VICENT, SERGIO; Mesquita, Amir Zacarías

    2011-01-01

    El uso de los códigos de Dinámica de Fluidos Computacional (CFD) 3D para predecir el flujo estacionario y transitorio en los reactores nucleares está creciendo rápidamente y constituye una nueva tendencia en su estudio. En este trabajo se presenta un modelo de un reactor de entrenamiento TRIGA (Training Research Isotopes General Atomic) tipo IPR-R1 desarrollado con el código de CFD ANSYS-CFX. El reactor nuclear TRIGA IPR-R1 está ubicado en el Centro de Desarrollo de Tecnología ...

  3. Resolution of the reactor vessel materials toughness safety issue; Task Action Plan A-11; Appendices C-K

    International Nuclear Information System (INIS)

    The central problem in the Unresolved Safety Issue A-11, 'Reactor Vessel Materials Toughness,' was to provide guidance in performing analyses for reactor pressure vessels (RPVs) which fail to meet the toughness requirements during service life as a result of neutron radiation embrittlement. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which has been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the RPV fracture problem with assumed beltline region flaws. Volume I of this report is a brief presentation of the problem and the results; Volume II provides the detailed technical foundations

  4. Practical application of the ''compact model'' for I and C modeling in the european pressurizer reactor (EPR) probabilistic safety assessment (PSA)

    International Nuclear Information System (INIS)

    The EPR is an advanced Pressurized Water Reactor (PWR) under development within FRAMATOME ANP and SIEMENS NPG, and with the support of their clients: a group of major German Utilities and the French Utility (EDF). Designers have expressed their willingness to engage probabilistic studies in the design process so as to assess the safety level in order to support and to optimize the choice of design options. This practice permits to reach well-balanced design and process, and to provide a reasonable assurance that the plant design is in compliance with the general safety objectives of the EPR. A level 1 Probabilistic Safety Assessment (PSA) has been set-up. The only topic of the level 1 PSA is the core damage assessment when the fuel elements are in place in the reactor vessel. Allowance is made for internal events only, including loss of off-site power or water supplies. The all plant conditions are considered from power states to reactor coolant system open to shutdown conditions. The EPR PSA is modeled with a computer tool which enables Boolean reduction and so takes into account all dependencies between systems as common pipes, power supplies, cooling and I and C. (author)

  5. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix C, marine transport and associated environmental impacts. Volume 2

    International Nuclear Information System (INIS)

    This is Appendix C to a Draft Environmental Statement on a Proposed Nuclear Weapon Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. Shipment of any material via ocean transport entails risks to both the ship's crew and the environment. The risks result directly from transportation-related accidents and, in the case of radioactive or other hazardous materials, also include exposure to the effects of the material itself. This appendix provides a description of the approach used to assess the risks associated with the transport of foreign research reactor spent nuclear fuel from a foreign port to a U.S. port(s) of entry. This appendix also includes a discussion of the shipping configuration of the foreign research reactor spent nuclear fuel, the possible types of vessels that could be used to make the shipments, the risk assessment methodology (addressing both incident-free and accident risks), and the results of the analyses. Analysis of activities in the port(s) is described in Appendix D. The incident-free and accident risk assessment results are presented in terms of the per shipment risk and total risks associated with the basic implementation of Management Alternative 1and other implementation alternatives. In addition, annual risks from incident-free transport are developed

  6. Radiation protection at the RA Reactor in 1986, Part -2, Annex 2c, Environmental Radioactivity control, meteorology measurements

    International Nuclear Information System (INIS)

    During 1986, meteorology measurements were continued as a part of the environmental control of the Vinca Institute. This report covers the period from November 1984 - November 1985. Part of the meteorology measurements and data analysis is adapted to the needs of the Institute, i.e. RA reactor and some Laboratories. The objective of these activities is forming the data base for solving everyday and special problems related to control, protection and safety of Institute environment

  7. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  8. High temperature mechanisms and kinetics of SiC oxidation under low partial pressures of oxygen: application to the fuel cladding of gas fast reactors

    International Nuclear Information System (INIS)

    Gas Fast Reactor (GFR) is one of the different Generation IV concepts under investigation for energy production. SiC/SiC composites are candidates of primary interest for a GFR fuel cladding use, thanks to good corrosion resistance among other properties. The mechanisms and kinetics of SiC oxidation under operating conditions have to be identified and quantified as the corrosion can decrease the mechanical properties of the composite. An experimental device has been developed to study the oxidation of silicon carbide under high temperature and low oxygen partial pressure. The results pointed out that not only parabolic oxidation, but also interfacial reactions and volatilization occur under such conditions. After determining the kinetics of each mechanism, as functions of oxygen partial pressure and temperature, the data are used for the modeling of the composites oxidation. The model will be used to predict the lifetime of the composite in operating conditions. (author)

  9. Hydrogen permeability, diffusivity, and solubility of SUS 316L stainless steel in the temperature range 400 to 800 .deg. C for fusion reactor applications

    International Nuclear Information System (INIS)

    Tritium permeation is one of the critical issues for the economy and safety of fusion power plants. As an initial step in tritium permeation research for fusion reactor applications, experiments were initiated by using hydrogen as a tritium substitute. An experimental system for hydrogen permeation and related behaviors in solid materials was designed and constructed. A continuous flow method was adopted with a capacity for high temperatures up to ∼1,000 .deg. C under ultra-high vacuums of ∼10-7 Pa. The hydrogen permeation behavior in SUS 316L stainless steel was investigated in the temperature range from 400 .deg. C to 800 .deg. C. As a result, the permeability, diffusivity and solubility of hydrogen were determined. The results were compared with the previously existing reference data. Changes in the sample's surface morphology after the hydrogen permeation experiment are also addressed.

  10. Refurbishing the reactor protection systems of VVER-440/230 and VVER-1000/320 nuclear power plants with exclusively digital IandC systems

    International Nuclear Information System (INIS)

    The refurbishment of reactor protection systems of nuclear power plants is based on two sets of requirements: engineering aspects such as performance, qualification and licensing, as well as interfaces to other systems; and cost-benefit relationships, ease of service and maintenance as well as installation during scheduled outages. A number of WWER-440 and WWER-1000 nuclear plants have announced their intention to refurbish their protection systems. Since 1994, these plants have been placing orders with Siemens for new protection systems, including the neutron flux monitoring system utilizing the advanced system TELEPERM XS. This exclusively digital IandC system provides an excellent foundation for the remaining plant service life

  11. Continuous-flow precipitation of hydroxyapatite at 37 °C in a meso oscillatory flow reactor

    OpenAIRE

    Castro, Filipa; Ferreira, António; Rocha, Fernando; A.A. Vicente; Teixeira, J. A.

    2013-01-01

    Continuous-flow precipitation of hydroxyapatite (HAp) was investigated in a meso oscillatory flow reactor (meso- OFR) and in a scaled-up meso-OFR, obtained by associating in series eight vertical meso-ORFs. Experiments were carried out under near-physiological conditions of temperature and pH, using fixed frequency ( f = 0.83 Hz) and amplitude (x0 = 4.5 mm), and varying the residence time from 0.4 to 6.7 min. It has been shown that the mean particle size and the aggregation degree of...

  12. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    International Nuclear Information System (INIS)

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging

  13. Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 k W. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 k W. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively. (author)

  14. Effect of carbon to nitrogen (C:N) ratio on nitrogen removal from shrimp production waste water using sequencing batch reactor.

    Science.gov (United States)

    Roy, Dhiriti; Hassan, Komi; Boopathy, Raj

    2010-10-01

    The United States Marine Shrimp Farming Program (USMSFP) introduced a new technology for shrimp farming called recirculating raceway system. This is a zero-water exchange system capable of producing high-density shrimp yields. However, this system produces wastewater characterized by high levels of ammonia, nitrite, and nitrate due to 40% protein diet for the shrimp at a high density of 1,000 shrimp per square meter. The high concentrations of nitrate and nitrite (greater than 25 ppm) are toxic to shrimp and cause high mortality. So treatment of this wastewater is imperative in order to make shrimp farming viable. One simple method of treating high-nitrogen wastewater is the use of a sequencing batch reactor (SBR). An SBR is a variation of the activated sludge process, which accomplishes many treatment events in a single reactor. Removal of ammonia and nitrate involved nitrification and denitrification reactions by operating the SBR aerobically and anaerobically in sequence. Initial SBR operation successfully removed ammonia, but nitrate concentrations were too high because of carbon limitation in the shrimp production wastewater. An optimization study revealed the optimum carbon to nitrogen (C:N) ratio of 10:1 for successful removal of all nitrogen species from the wastewater. The SBR operated with a C:N ratio of 10:1 with the addition of molasses as carbon source successfully removed 99% of ammonia, nitrate, and nitrite from the shrimp aquaculture wastewater within 9 days of operation. PMID:20835881

  15. A method for assessing the annual dose to the most exposed individual from tritium and 14C reactor discharges to atmosphere

    International Nuclear Information System (INIS)

    A method is described for assessing the annual dose to the most exposed individual from routine releases of tritium and 14C to the atmosphere during normal reactor operations. A detailed assessment has been made of the resulting equilibrium contamination levels in a range of foodstuffs typical of an average UK diet and of the annual doses resulting from a chronic intake of tritium and 14C via inhalation, ingestion and, additionally, in the case of tritium, via skin absorption. Equilibrium annual doses from the global circulation of tritium and 14C have also been calculated. Upper limits to the effective annual dose-equivalents to the most exposed individual were found to be 0.6 rem.yr-1 and 100 rem.yr-1 per Ci.yr-1 release of tritium and 14C respectively, with the ingestion pathway contributing significantly to the overall exposure. The most exposed individual was found to be a Reference 10 year old child. The methods outlined for calculating the ingestion dose from tritium and 14C releases hav been incorporated into the more generally applicable code FOODDOSE. The code may be used to make more realistic dose calculations to the individuals based on site-specific surveys of variables such as local meteorology, local diet and local land use for agriculture, which may lead to doses smaller than the upper limit values quoted by factors of 20 and 200 for tritium and 14C respectively. (author)

  16. Effects of post-reactor functionalization on the phase behaviour of an ethylene-1-octene copolymer studied using solid-state high resolution 13C NMR spectroscopy.

    Science.gov (United States)

    Calucci, Lucia; Cicogna, Francesca; Forte, Claudia

    2013-10-01

    The effects of post-reactor functionalization with naphthoate-TEMPO on the structure and morphology of an ethylene-1-octene copolymer were investigated by means of solid-state NMR techniques and DSC measurements. Selective (13)C MAS experiments allowed the orthorhombic and the monoclinic crystalline phases and two amorphous phases with different degree of mobility to be detected and quantified. (13)C and (1)H relaxation time measurements and spin diffusion experiments gave insight into the polymer dynamics within the different phases, the crystalline domain dimensions, and the rate of chain diffusion between amorphous and crystalline phases. Comparison of the results obtained for the pristine copolymer and the functionalized samples clearly indicated that the functionalization procedure causes redistribution within the crystalline and the amorphous phases with no relevant change in the degree of crystallinity or in the crystalline domain average size, and slows down chain diffusion. PMID:23942957

  17. Activity evaluation of 60Co, 137Cs, 152Eu, 154Eu, 14C and total β in graphite from the thermal column of Romanian TRIGA research reactor

    International Nuclear Information System (INIS)

    Several methods were investigated for identification and activity evaluation of 60Co, 137Cs, 152Eu, 154Eu, 14C and total β in graphite samples collected from the thermal column of Romanian research reactor. An ORTEC DIGIDART multichannel gamma spectrometer was used for measurement of 60Co, 137Cs, 152Eu and 154Eu. The main contribution to the radioactivity of graphite comes from 60Co, 152Eu and 154Eu. Combustion of graphite samples using a PerkinElmer sample oxidizer was utilized for total oxidizing of 14C to gaseous carbon dioxide. A TRI-CARB Packard liquid scintillation analyzer was used for 14C counting. The degree of 14C recovery in the analyzed standard samples was higher than 98%, and the 14C retention degree in the circuits of combustion system was less than 0.08%. To evaluation of total β activity in the irradiated graphite, the samples were measured using a proportional counter. The mean total β activity in the irradiated graphite samples was about 3000 Bq/g. (authors)

  18. Reactor Physics

    International Nuclear Information System (INIS)

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

  19. Reactor Physics

    International Nuclear Information System (INIS)

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

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  1. On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Kwon; Kramer, Daniel [Center for Electrochemical Science and Technology, Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Macdonald, Digby D., E-mail: macdonald@berkeley.edu [Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720 (United States)

    2014-11-15

    Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288 °C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji’s model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length.

  2. On the shape of stress corrosion cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant piping at 288 °C

    International Nuclear Information System (INIS)

    Evolution of the shape of surface cracks in sensitized Type 304 SS in Boiling Water Reactor primary coolant circuit piping at the reactor operating temperature of 288 °C is explored as a function of various environmental variables, such as electrochemical potential (ECP), solution conductivity, flow velocity, and multiplier for the oxygen reduction reaction (ORR) standard exchange current density (SECD), using the coupled environment fracture model (CEFM). For this work, the CEFM was upgraded by incorporating Shoji’s model for calculating the crack tip strain rate and more advanced expressions were used for estimating the stress intensity factor for semi-elliptical surface cracks. This revised CEFM accurately predicts the dependence of the crack growth rate on stress intensity factor and offers an alternative explanation for the development of semi-elliptical cracks than that provided by fracture mechanics alone. The evolution of surface crack semi-elliptical shape depends strongly upon various environmental variables identified above, and the CEFM predicts that the minor axis of the ellipse should be oriented perpendicular to the surface, in agreement with observation. The development of the observed semi-elliptical cracks with the minor axis perpendicular to the surface is therefore attributed to the dependence of the crack growth rate on the electrochemical crack length

  3. Determination of Carbon-14 (14C) and Tritium (3H) in the graphite thermal column of a IRT-2000 research reactor in Sofia

    International Nuclear Information System (INIS)

    Radiocarbon has average, and tritium has low level of radiotoxicity, but both are easily absorbable radionuclide from the body as due to their chemical nature they are included in almost all the chemical compounds involved in metabolic chains and in structure of proteins and biological macromolecules. Due to above mentioned reasons, the analysis and determination of these two major radionuclides is very significant for radiation protection for both the public and the environment as well as for occupational exposure of personnel. This publication describes the detailed procedure for determining the concentration of 3H and 14C in samples selected at the stage 'partial dismantling' of the research nuclear reactor IRT-2000 reconstruction. The procedure consists graphite sample thermal burning in the tubing muffle furnace at a maximum temperature of 850'0C. After combustion the gas mixture passes through a CuO catalyst, heated to 450-5000C and through a series of traps for 3H and 14C capturing. The captured CO2+ 14CO2 as Na2CO3 in NaOH precipitates with BaCl2 as BaCO3 and it is followed by a liquid scintillation spectrometry measurements

  4. Impact of increasing the enrichment of 15N in reduced-activation ferritic steels on 14C production of tokamak reactors

    International Nuclear Information System (INIS)

    The impact of increasing the enrichment of 15N in F82H of a fusion reactor has been investigated in order to increase the fraction of shallow land burial. In Japan, the radioactive waste having any single radionuclide, exceeding the limit concentration value determined by the Nuclear Safety Commission, will not qualify as a low level waste (LLW), which can be disposed by shallow land burial. The regulation of carbon-14, which has a half life of 5730 years, was 37 kBq/g. Low activation ferritic steel is the most promising structural materials under research for fusion because of its good irradiation properties at high temperature and low activation. From the viewpoint of toughness and reliability, it is preferable to mix more than 200 ppm nitrogen in F82H. In DT fusion neutron environment, 14C is produced by 14N(n,p)14C nuclear reaction. The natural abundance of nitrogen isotopes is 99.63% of 14N and 0.37% of 15N. Then increasing 15N enrichment is effective to reduce the production of 14C. SlimCS, which is a compact demonstration reactor characterized by low aspect ratio (A) and reduced-size center solenoid (CS), has been designed in JAEA. The reactor has a major radius of 5.5 m and aspect ratio of 2.6. It produces a fusion output of 2.95 GW, which corresponds to a neutron wall loading of 3.6 MW/m2. A tritium breeding blanket in outboard side consists of both 30 cm thick replaceable blanket and 50 cm thick permanent blanket. In the inboard side, it has only replaceable blanket of 30 cm thickness to realize low-A. Neutron transport calculations have been performed using ANISN with nuclear cross sections from the FUSION-40 based on JENDL 3.1. The concentration of nitrogen in F82H is 200 ppm. The enrichment of 15N was varied from natural abundance of 0.37% to 95%. After thirty years operation with the operating rate of 50%, the 14C concentrations in F82H of the surface of outboard permanent blanket decreased from 78 kBq/g to 10 kBq/g by increasing the enrichment of 15N

  5. Study of the hydrogen behavior in amorphous hydrogenated materials of type a - C:H and a - SiC:H facing fusion reactor plasma

    International Nuclear Information System (INIS)

    Plasma facing components of controlled fusion test devices (tokamaks) are submitted to several constraints (irradiation, high temperatures). The erosion (physical sputtering and chemical erosion) and the hydrogen recycling (retention and desorption) of these materials influence many plasma parameters and thus affect drastically the tokamak running. First, we will describe the different plasma-material interactions. It will be pointed out, how erosion and hydrogen recycling are strongly related to both chemical and physical properties of the material. In order to reduce these interactions, we have selected two amorphous hydrogenated materials (a-C:H and a-SiC:H), which are known for their good thermal and chemical qualities. Some samples have been then implanted with lithium ions at different fluences. Our materials have been then irradiated with deuterium ions at low energy. From our results, it is shown that both the lithium implantation and the use of an a - SiC:H substrate can be beneficial in enhancing the hydrogen retention. These results were completed with thermal desorption studies of these materials. It was evidenced that the hydrogen fixation was more efficient in a-SiC:H than in a-C:H substrate. Results in good agreement with those described above have been obtained by exposing a - C:H and a - SiC:H samples to the scrape off layer of the tokamak of Varennes (TdeV, Canada). A modelling of hydrogen diffusion under irradiation has been also proposed. (author)

  6. Hydrogen behaviour study in plasma facing a-C:H and a-SiC:H hydrogenated amorphous materials for fusion reactors

    International Nuclear Information System (INIS)

    Plasma facing components of controlled fusion test devices (tokamaks) are submitted to several constraints (irradiation, high temperatures). The erosion (physical sputtering and chemical erosion) and the hydrogen recycling (retention and desorption) of these materials influence many plasma parameters and thus affect drastically the tokamak running. Firstly, we will describe the different plasma-material interactions. It will be pointed out, how erosion and hydrogen recycling are strongly related to both chemical and physical properties of the material. In order to reduce this interactions, we have selected two amorphous hydrogenated materials (a-C:H and a-SiC:H), which are known for their good thermal and chemical qualities. Some samples have been then implanted with lithium ions at different fluences. Our materials have been then irradiated with deuterium ions at low energy. From our results, it is shown that both the lithium implantation and the use of an a-SiC:H substrate can be benefit in enhancing the hydrogen retention. These results were completed with thermal desorption studies of these materials. It was evidenced that the hydrogen fixation was more efficient in a -SiC:H than in a-C:H substrate. Results in good agreement with those described above have been obtained by exposing a-C:H and a-SiC:H samples to the scrape off layer of the tokamak of Varennes (TdeV, Canada). A modeling of hydrogen diffusion under irradiation has been also proposed. (author)

  7. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  8. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

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

  9. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 6, appendices A, B, and C

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events (including internal flooding, but excluding internal fire). The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, reviewed the WE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. In particular, these results are assessed in relation to the design and operational characteristics of the various reactor and containment types, and by comparing the IPEs to probabilistic risk assessment characteristics. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants.

  10. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 6, appendices A, B, and C

    International Nuclear Information System (INIS)

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events (including internal flooding, but excluding internal fire). The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, reviewed the WE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. In particular, these results are assessed in relation to the design and operational characteristics of the various reactor and containment types, and by comparing the IPEs to probabilistic risk assessment characteristics. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants

  11. Modeling of mechanical behavior and design criteria for SiC{sub f}/SiC composite structures in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G.; Giancarli, L. E-mail: luciano.giancarli@cea.fr; Golfier, H.; Maire, J.-F

    2003-01-01

    Silicon carbide composites are the primary composite materials being evaluated and developed world-wide for fusion structural applications. Their use as structural material involves however the adoption of appropriate design methodologies capable to take into account their complex non linear mechanical behavior. Some behavioral models developed specifically for SiC{sub f}/SiC composites and already available in literature are reviewed in this paper. They have been evaluated with respect to the possibility of implementation in FEM codes and keeping in mind fusion-specific issues. One of them has recently been implemented in the FEM code CASTEM 2000. At the same time, appropriate resistance criteria are necessary to correctly assess the potential advantages of SiC{sub f}/SiC composites over more common metallic materials. A new resistance criterion has then been defined and is presented in this work. Examples of application of the improved design analysis methodology are given throughout the paper.

  12. Stress corrosion cracking of reactor pressure vessel steel in 288 C water: The effect of oxygen, electrochemical potential, and steel composition

    International Nuclear Information System (INIS)

    A new set of fracture mechanics stress corrosion crack growth rate data is presented for transgranular cracking of reactor pressure steels in high temperature water. The essential observations are as follows. Fast stress corrosion crack growth rates between 10-9 and 10-8 m/s may be observed down to 400 ppb dissolved oxygen at water conductivities of 1.0 μs/cm in refreshed autoclaves. Stress corrosion crack growth could not be observed below minus 270 mV on the hydrogen scale. Low and medium sulfur contents in the steels have no measurable influence on the stress corrosion crack growth rates in 288 C water with a conductivity of 1.0 μS/cm

  13. Processing and characterization of B4C/Cu graded composite as plasma facing component for fusion reactors

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    A new approach for fabricating B4C/Cu graded composite by rapid self-resistance sintering under ultra-high pressure was presented, by which a near dense B4C/Cu graded composite with a compositional spectrum of 0-100% was successfully fabricated. Plasma relevant performances ofsintered B4C/Cu composite were preliminarily characterized, it is found that its chemical sputtering yield is 70% lower than that of SMF800 nuclear graphite under 2.7 keV D+ irradiation, and almost no damages after 66 shots of in situ plasma discharge in HL-1 Tokamak facility, which indicates B4C/Cu plasma facing component has a good physical and chemical sputtering resistance performance compared with nuclear graphite.

  14. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  15. Determination of the fractions of syntrophically oxidized acetate in a mesophilic methanogenic reactor through an (12)C and (13)C isotope-based kinetic model.

    Science.gov (United States)

    Gehring, Tito; Niedermayr, Andrea; Berzio, Stephan; Immenhauser, Adrian; Wichern, Marc; Lübken, Manfred

    2016-10-01

    In order to accurately describe the carbon flow in anaerobic digestion processes, this work investigates the acetate degradation pathways through the use of stable carbon isotope analysis and a mathematical model. Batch assays using labeled (13)C acetate were employed to distinguish the acetate consumption through methanogenic Archaea and acetate-oxidizing Bacteria. Suspended and sessile biomass, with over 400 days of retention time, from a mesophilic (36.5 °C) upflow anaerobic filter was used as inocula in these assays. A three-process model for acetoclastic methanogenesis and syntrophic acetate oxidation (SAO) was developed to allow for a precise quantification of the SAO contribution. The model distinguishes carbon atoms in light and heavy isotopes, (12)C and (13)C, respectively, which permitted the simulation of the isotope ratios variation in addition to gas production, gas composition and acetate concentrations. The model indicated oxidized fractions of acetate between 7 and 18%. Due to the low free ammonia inhibition potential for the acetoclastic methanogens in these assays these findings point to the biomass retention times as a driven factor for the SAO pathway. The isotope-based kinetic model developed here also describes the δ(13)C variations in unlabeled assays accurately and has the potential to determine biological (13)C fractionation factors. PMID:27390036

  16. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  17. Ti{sub 3}SiC{sub 2} as a candidate material for lead cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Utili, M., E-mail: marco.utili@enea.i [Universita degli studi di Bologna, Facolta di Ingegneria, Via Risorgimento, 2 DIENCA, Bologna (Italy); Agostini, M.; Coccoluto, G. [ENEA CR Brasimone, Fis Ing, Camugnano (Italy); Lorenzini, E. [Universita degli studi di Bologna, Facolta di Ingegneria, Via Risorgimento, 2 DIENCA, Bologna (Italy)

    2011-05-15

    One of the main issues regarding the design of LFR is the corrosion-erosion behaviour of materials in contact with high temperature and high velocity lead, such as pump impeller and bearing materials. Those materials will worked in lead at about 480 {sup o}C with a velocity relative to impeller blade in the magnitude of 10 m/s. Ti{sub 3}SiC{sub 2}, a ternary compound was selected as one of the most promising candidate material, the most significant aspect of Ti{sub 3}SiC{sub 2} is that it combines some of the most attractive proprieties of ceramics with those of metals. This material was already applied in industry but more effort is needed to qualify its performance when compared with candidate structural stainless steel available in Europe. With the purpose of evaluating the corrosion resistance of Ti{sub 3}SiC{sub 2}, a preliminary screening test was carried out at ENEA Brasimone in fluent lead under oxidation conditions (oxygen concentration of about 10{sup -6} wt% in the melt), with an average temperature of 500 {sup o}C and a stream velocity of about 1 m/s.

  18. High-effective denitrification of low C/N wastewater by combined constructed wetland and biofilm-electrode reactor (CW-BER).

    Science.gov (United States)

    He, Yuan; Wang, Yuhui; Song, Xinshan

    2016-03-01

    The low denitrification effect on constructed wetlands (CWs) treating low carbon to nitrogen ratio (C/N) wastewater was a problem. In this study, a novel coupled system by installing CW and biofilm-electrode reactor (CW-BER) was developed. In this system, the heterotrophic and autotrophic denitrifying bacteria all played their roles in denitrification process. The system was investigated systematically with simulated wastewater at different C/Ns, electric current intensities (I), hydraulic retention times (HRTs), and pH. Results showed that the optimum running conditions were C/N=0.75-1, I=15 mA, HRT=12 h, and pH=7.5. The highest removal efficiency of NO3-N and TN at the best conditions was respectively 63.03% and 98.11% for CW-BER. Also, the TN and NO3-N enhancive removal efficiency of CW-BER was 23.26% and 24.20%, respectively. No residual organic carbon source was detected in final effluent at the best parameters. PMID:26735879

  19. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    International Nuclear Information System (INIS)

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS; Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  20. Measurement of the enthalpy and specific heat of a Be2C-graphite-UC2 reactor fuel material to 19800K

    International Nuclear Information System (INIS)

    The enthalpy and specific heat of a Be2C-graphite-UC2 composite nuclear fuel material were measured over the temperature range 300 to 19800K using differential scanning calorimetry and liquid argon vaporization calorimetry. The fuel material measured was developed at Sandia National Laboratories for use in pulsed test reactors. The material is a hot-pressed composite consisting of 40 vol % Be2C, 49.5 vol % graphite, 3.5 vol % UC2 and 7.0 vol % void. The specific heat was measured with the differential scanning calorimeter over the temperature range 300 to 9500K while the enthalpy was measured over the range 1185 to 19800K with the liquid argon vaporization calorimeter. The normal spectral emittance at a wavelength of 6.5 x 10-5 cm was measured over the experimental temperature range. The combined experimental enthalpy data were fit using a spline routine and differentiated to give the specific heat. Comparison of the measured specific heat of the composite to the specific heat calculated by summing the contributions of the individual components indicates that the specific heat of the Be2C component differs significantly from literature values and is approximately 0.6 cal/g-K (2.5 x 103 J/Kg-K) for temperatures above 10000K

  1. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS); Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  2. Perfeccionamiento del modelo de vasija del reactor de la central de Ascó para el código de cálculo Relap5

    OpenAIRE

    Berggren Durall, Alberto

    2012-01-01

    El objetivo principal del proyecto realizado es el de perfeccionar el modelo de vasija de la central nuclear de Ascó utilizando el código de cálculo RELAP5/mod3.3. Con este fin se proponen y estudian tres variaciones en la modelación de la vasija perteneciente al modelo general de planta y se escogen dos escenarios relevantes para analizar y comparar los comportamientos. En primer lugar, a partir de un cambio real efectuado en la CNA se desarrolla una nueva nodalización de l...

  3. Development of metal-carbon eutectic cells for application as high temperature reference points in nuclear reactor severe accident tests: Results on the Fe-C, Co-C, Ti-C and Ru-C alloys' melting/freezing transformation temperature under electromagnetic induction heating

    International Nuclear Information System (INIS)

    With the aim of reducing the high temperature measurement uncertainty of nuclear reactor severe accident experimental tests at the PLINIUS platform in Cadarache Research Centre, France, a variety of graphite cells containing a metal-carbon eutectic mix have been tested to assess the melting/freezing temperature reproducibility and their feasibility as calibration cells for thermometers. The eutectic cells have been thermally cycled in an induction furnace to assess the effect of heating/cooling rate, metal purity, graphite crucible design, and binary system constituents on the eutectic transformation temperature. A bi-chromatic pyrometer was used to perform temperature measurements in the graphite cell black cavity containing the metal-carbon eutectic mix. The eutectic points analyzed are all over 1100 C and cover an almost thousand degree span, i.e. from the Fe-Fe3C to the Ru-C eutectic. The induction heating permitted the attainment of heating and cooling rates of over 200 C/min under an inert atmosphere. The conducted tests allowed the determination of general trends and peculiarities of the solid. liquid transformation temperature under non-equilibrium and non-steady-state conditions of a variety of eutectic alloys (Fe-C, Co-C, Ti-C and Ru-C binary systems). (authors)

  4. Research reactors

    International Nuclear Information System (INIS)

    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)

  5. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

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

  6. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  7. Reactor container

    International Nuclear Information System (INIS)

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  8. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  9. Trace metal assay of fast breeder test reactor fuel using D.C. arc and plasma emission spectrometry

    International Nuclear Information System (INIS)

    This report describes the methods developed and used for the trace metal assay of the first charge of the FBTR fuel using a glove-box adapted direct reading emission spectrometer. The group of medium and highly volatile elements are determined in (U,Pu)C fuel samples by d.c. arc carrier distillation technique while the group of lanthanide elements are determined by ICP excitation mode with prior chemical separation from the major matrix. The statistical treatment of the analytical data collected from the analysis of about one hundred samples has indicated good purity of samples and consistent and satisfactory performance of the direct reading spectrometer and associated systems during this period. (author)

  10. Reactor production of n.c.a grade 143Pr, 161Tb and 111Ag radionuclides for radionuclide therapy applications

    International Nuclear Information System (INIS)

    Large scale radioisotope production of no-carrier-added (n.c.a) radionuclides from bulk amounts of neutron irradiated targets were evaluated in view of their potential applications in nuclear medicine. Three different radionuclides viz 143Pr (T1/2 13.57 d and Eβmax 0.937 MeV), 161Tb (t1/2 6.91 d, Eβmax 593.1keV, Eγ 74.6 keV (9.8%)) and 111Ag (t1/2 7.45 d, Eβmax 1.09 MeV, Eγ 245.4 keV (1.24%), 342.1 KeV (6.68%)) were investigated for production in n.c.a form by (n,γ) followed by beta decay route. The radiochemical separation of the nuclide of interest was performed by precipitation in case of 143Pr and 161Tb and an ion exchange chromatography for 111Ag. The separated n.c.a grade products were of high radionuclidic purity (>99.9%) and did not contain any detectable impurities. Logistics of regular production of these radionuclides, their quality control and the scope for production of these radiochemicals with high specific activity are discussed in this paper. (author)

  11. Several loadings and stresses of first wall of SiC with metal liner on conceptual design of moving ring reactor 'KARIN-1'

    International Nuclear Information System (INIS)

    On conceptual design of moving ring reactor ''KARIN-I'' (Output: 1850 MWe), the first wall of SiC with metal liner is considered by reason that SiC ceramics has specific features of excellent radiation damage resistance in fast neutron spectra and a very low residual radioactivity, and that the thin metal liner has good compatibility with liquid lithium and good vaccum-tight, however, a extent electromagnetic interaction. The electromagnetic force applied on the metal liner and several pressure losses of liquid lithum flow are estimated, and these forces correspond to the fluid mechanical loading on SiC first wall. Thermal loading by neutron flux is calculated on the first wall to obtain temperature distributions along the flow direction and toward the wall thickness. At the outlet of the burning section, the surface temperature of SiC rises to the value of 8250C on plasma side and on the metal liner, it rises to the value of 5400C. Finally, the stress analysis is performed. The thermal stress is about one order larger than the stress induced by the fluid mechanical loading. At the inlet of the burning section, the average tensile stress of 22.4kg/mm2 is induced on the outer side of SiC wall, and on the inner side, the average compressive stress of -26.1kg/mm2 is induced. At the outlet of the burning section, the tensile stress is found to oscillate between 25.5kg/mm2 and 27.3kg/mm2 on the outer side of SiC wall by frequency of 1 Hz, and on the inner side, the compressive stress also oscillates between -21.6kg/mm2 and -29.0kg/mm2 by the same frequency. These stresses are within the value of fracture stress, (72.5kg/mm2). Difficult residual problems on the first wall are also discussed. (author)

  12. Graphene-modified Pd/C cathode and Pd/GAC particles for enhanced electrocatalytic removal of bromate in a continuous three-dimensional electrochemical reactor.

    Science.gov (United States)

    Mao, Ran; Zhao, Xu; Lan, Huachun; Liu, Huijuan; Qu, Jiuhui

    2015-06-15

    Bromate (BrO3(-)) is a carcinogenic and genotoxic contaminant commonly generated during ozonation of bromide-containing water. In this work, the reductive removal of BrO3(-) in a continuous three-dimensional electrochemical reactor with palladium-reduced graphene oxide modified carbon paper (Pd-rGO/C) cathode and Pd-rGO modified granular activated carbon (Pd-rGO/GAC) particles was investigated. The results indicated that the rGO sheets significantly promoted the electrochemical reduction of BrO3(-). With the enhanced electron transfer by rGO sheets, the electroreduction of H2O to atomic H* on the polarized Pd particles could be significantly accelerated, leading to a faster reaction rate of BrO3(-) with atomic H*. The synergistic effect of the Pd-rGO/C cathode and Pd-rGO/GAC particles were also exhibited. The atomic H* involved in various electroreduction processes was detected by electron spin resonance spectroscopy and its role for BrO3(-) reduction was determined. The performance of the reactor was evaluated in terms of the removal of BrO3(-) and the yield of Br(-) as a function of the GO concentration, Pd loading amount, current density, hydraulic residence time (HRT), and initial BrO3(-) concentration. Under the current density of 0.9 mA/cm(2), BrO3(-) with the initial concentration of 20 μg/L was reduced to be less than 6.6 μg/L at the HRT of 20 min. The BrO3(-) reduction was inhibited in the presence of dissolved organic matter. Although the precipitates generated from Ca(2+) and Mg(2+) in the tap water would cover the Pd catalysts, a long-lasting electrocatalytic activity could be maintained for the 30 d treatment. SEM and XPS analysis demonstrated that the precipitates were predominantly deposited onto the Pd-rGO/C cathode rather than the Pd-rGO/GAC particles. PMID:25834955

  13. Optimization of C/N Ratio and Inducers for Wastewater Paper Industry Treatment Using Trametes versicolor Immobilized in Bubble Column Reactor

    Directory of Open Access Journals (Sweden)

    Aura M. Pedroza-Rodríguez

    2013-01-01

    Full Text Available C/N ratio and MnSO4 and CuSO4 concentrations were optimized for decolorization and chemical oxygen demand (COD removal of bleached Kraft pulp mill effluent by Trametes versicolor immobilized in polyurethane foam. Statistical differences (P<0.0001 at high C/N ratios (169, 2 mM CuSO4, and 0.071 mM MnSO4 were determined. Decolorization of 60.5%, COD removal of 55%, laccase (LAC 60 U/L, and manganese peroxidase (MnP 8.4 U/L were obtained. Maximum of decolorization (82%, COD removal (83%, LAC (443.5 U/L, and MnP (18 U/L activities at C/N ratio of 405 (6.75 mM CuSO4 and 0.22 mM MnSO4 was achieved in step 7 at 4 d. Positive correlation between the decolorization, COD removal, and enzymatic activity was found (P<0.0001. T. versicolor bioremediation capacity was evaluated in bubble column reactor during 8 d. Effluent was adjusted according to optimized parameters and treated at 25°C and air flow of 800 mL/min. Heterotrophic bacteria growth was not inhibited by fungus. After 4 d, 82% of COD reduction and 80% decolorization were recorded. Additionally, enzymatic activity of LAC (345 U/L and MnP (78 U/L was observed. The COD reduction and decolorization correlated positively (P<0.0001 with enzymatic activity. Chlorophenol removal was 98% of pentachlorophenol (PCP, 92% of 2,4,5-trichlorophenol (2,4,5-TCP, 90% of 3,4-dichlorophenol (3,4-DCP, and 99% of 4-chlorophenols (4CP.

  14. Low cycle fatigue behaviour of neutron irradiated copper alloys at 250 and 350 deg. C. (ITER R and D Task no. T213)[International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B.N.; Stubbins, J.F. [Illinois Univ., Dept. of Nuclear Engineering, Illinois (United States); Toft, P

    2000-03-15

    The fatigue behaviour of a dispersion strengthened and a precipitation hardened copper alloys was investigated with and without irradiation exposure. Fatigue specimens of these alloys were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of {approx}2.5 x 10{sup 17} n/m{sup 2}s (E> 1 MeV) to influence levels of 1.0 - 1.5 x 10{sup 24} n/m{sup 2} (E> 1 MeV) at 250 and 350 deg. C. These irradiations were carried out in temperature controlled rigs where the irradiation temperature was monitored and controlled continuously throughout the whole irradiation experiment. Both unirradiated and irradiated specimens were fatigue tested in vacuum at the irradiation temperatures of 250 and 350 deg. C in a strain controlled mode with a loading frequency of 0.5Hz. Post-fatigue microstructures were examined using transmission electron microscopy and the fracture surfaces were investigated using scanning electron microscope. The present investigations demonstrated that the fatigue life decreases with increasing temperature and that the exposure to neutron irradiation causes further degradation in fatigue life at both temperatures. These results are discussed in terms of the observed post-fatigue microstructures and the fracture surface morphology. Finally, the main conclusions and their implications are summarised. (au)

  15. Greater-than-Class C low-level waste characterization. Appendix G: Evaluation of potential for greater-than-Class C classification of irradiated hardware generated by utility-operated reactors

    International Nuclear Information System (INIS)

    This study compiles and evaluates data from many sources to expand a base of data from which to estimate the activity concentrations and volumes of greater-than-Class C low-level waste that the Department of Energy will receive from the commercial power industry. Sources of these data include measurements of irradiated hardware made by or for the utilities that was classified for disposal in commercial burial sites, measurements of neutron flux in the appropriate regions of the reactor pressure vessel, analyses of elemental constituents of the particular structural material used for the components, and the activation analysis calculations done for hardware. Evaluations include results and assumptions in the activation analyses. Sections of this report and the appendices present interpretation of data and the classification definitions and requirements

  16. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  17. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  18. Heterogeneous reactors

    International Nuclear Information System (INIS)

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author)

  19. Physical security at research reactors

    International Nuclear Information System (INIS)

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  20. Reactivity determination of the Al2O3-B4C burnable poison as a function of its concentration in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Burnable poison rods made of Al2O3-B4C pellets with different concentrations of 10B have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. The experiments evaluated the reactivity of the burnable poison rods as a function of the 10B concentration, and the shadowing effect on the control rod reactivity worth as a function of the distance between the burnable position rods and the control rod. The results showed that the burnable poison rods have a non-linear behavior as function of the 10 B concentration, starting to reach an asymptotic value for concentrations higher than 7 g/cm3 of 10B. The shadowing effect on the control rods was substantial. When the burnable poison rods were beside the control rod, its reactivity worth decreased as much as 30 %, and when they were 10,5 cm distant, the control rod worth decreased by 7 %. The MCNP results for the burnable poison reactivity effects agreed within experimental errors with the measured values. (author)

  1. Evaluation of liquefaction temperature of metallic fast reactor fuel. Study on the U-Pu-Fe phase diagram at temperatures around 650degC

    International Nuclear Information System (INIS)

    Since metallic fuel fast reactors have a large potential for safety, economy, and fuel-multiply, Central Research Institute of Electric Power Industry has been developed the metallic fuel. A peak cladding inner temperature of the metallic fuel should be lower than the temperature at which a liquid phase is formed in the reaction zone at the fuel alloy-cladding interface during the normal operation. In the present study, a ternary phase diagram for the U-Pu-Fe system, at which the liquefaction temperature is considered to become the lowest, was determined using a method of combination of thermodynamic calculation and experimental study. the results indicated that a liquefaction temperature can be estimated from the Pu solubility limit in FeU6 compound. The Pu solubility limit in FeU6 compound attained 14% at 650C and, therefore, no liquefaction was predicted to occur in the interaction layer between fuel and cladding, in the condition that the Pu concentration in FeU6 compound is maintained to be lower than 14%. (author)

  2. Ammonia oxidizing bacteria and archaea in horizontal flow biofilm reactors treating ammonia-contaminated air at 10 °C.

    Science.gov (United States)

    Gerrity, Seán; Clifford, Eoghan; Kennelly, Colm; Collins, Gavin

    2016-05-01

    The objective of this study was to demonstrate the feasibility of novel, Horizontal Flow Biofilm Reactor (HFBR) technology for the treatment of ammonia (NH3)-contaminated airstreams. Three laboratory-scale HFBRs were used for remediation of an NH3-containing airstream at 10 °C during a 90-d trial to test the efficacy of low-temperature treatment. Average ammonia removal efficiencies of 99.7 % were achieved at maximum loading rates of 4.8 g NH3 m(3) h(-1). Biological nitrification of ammonia to nitrite (NO2 (-)) and nitrate (NO3 (-)) was mediated by nitrifying bacterial and archaeal biofilm populations. Ammonia-oxidising bacteria (AOB) were significantly more abundant than ammonia-oxidising archaea (AOA) vertically at each of seven sampling zones along the vertical HFBRs. Nitrosomonas and Nitrosospira, were the two most dominant bacterial genera detected in the HFBRs, while an uncultured archaeal clone dominated the AOA community. The bacterial community composition across the three HFBRs was highly conserved, although variations occurred between HFBR zones and were driven by physicochemical variables. The study demonstrates the feasibility of HFBRs for the treatment of ammonia-contaminated airstreams at low temperatures; identifies key nitrifying microorganisms driving the removal process; and provides insights for process optimisation and control. The findings are significant for industrial applications of gas oxidation technology in temperate climates. PMID:26879980

  3. Plasma reactor

    OpenAIRE

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  4. Reactor physics

    International Nuclear Information System (INIS)

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  5. 25 years operating experience with the Regional Centre for Nuclear Research of Kinshasa (C.R.E.N.-K) TRIGA reactor

    International Nuclear Information System (INIS)

    Reactor operation began in Zaire with the start up of the 50 KW TRIGA Mark I reactor in 1959: the very first reactor in Africa. After eleven years of operation the TRIGA Mark I reactor facility was shut down in June 1970, dismantled later on while construction began for the TRIGA Mark II version. The new reactor was loaded to critically on March 24, 1972. It is a TRIGA Mark II, F-ring, graphite reflected reactor was a steady-state power level of 1 MW and pulsing capability of up to 1500 MW. Thermal neutron flux at the power level of 1 MW equals about 1013 n/cm2.s in the central thimble. The 20 % enriched U-Zr-H1.6 fuel is contained in stainless steel cladded elements. The core loading contains 66 type 104 standard fuel elements, 1 type 204 instrumented fuel element and 3 type 304 fueled follower control rods (FFCR'S). Below, we present some important figures related to routine utilization and problems met during about 14 years of operation of the TRIGA Mark II reactor. (author)

  6. Caracterización de los parámetros neutrónicos del reactor Triga IPR-R1 con el código SCALE6.0 (KENO VI)

    OpenAIRE

    FARIA DE CASTRO, VICTOR; Miró Herrero, Rafael; Mello da Silva, C. A.; Pereira, C; Verdú Martín, Gumersindo Jesús; BARRACHINA CELDA, TERESA MARÍA; Dalle, H. M.

    2011-01-01

    KENO - VI es un código de transporte neutrónico en el sistema SCALE6.0 que utiliza el método de Monte Carlo para cálculos de criticidad en sistemas nucleares. Se ha construido un modelo en 3D para caracterizar parámetros neutrónicos del reactor de investigación TRIGA IPR - R1. Se han comparado los valores encontrados con los obtenidos por el código MCNP5 y experimentales con el propósito de validar esta metodología.

  7. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  8. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  9. Draft environmental impact statement for the siting, construction, and operation of New Production Reactor capacity. Volume 3, Sections 7-12, Appendices A-C

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

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

  10. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  11. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  12. 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.

  13. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  14. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  15. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  17. Breeder reactors

    International Nuclear Information System (INIS)

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

  18. Effect of dynamic strain aging on the low cycle fatigue in SA508 C1.3 reactor pressure vessel steel

    International Nuclear Information System (INIS)

    The effect of dynamic strain aging on cyclic stress response and fatigue life of ASME SA508 Cl.3 forging steel for nuclear reactor pressure vessels was studied at temperatures ranging from room temperature to 500 .deg. C. Total strain ranges and strain rates were varied from 0.7 to 2.0% and from 1x10-2 to 4x10-4s-1, respectively. Test environments were the atmospheres of air and argon. Except for the temperature region of dynamic strain aging, 300 .deg. C, the initial cyclic hardening was immediately followed by cyclic softening at all strain rates in air atmosphere. The cyclic softening continued over 95% of the fatigue life. On the other hand, at the dynamic strain aging temperature, the operating temperature of nuclear pressure vessel, the variation of cyclic stress amplitude showed the primary and secondary hardening dependent on the strain rate and the total strain range. Dynamic strain aging was manifested as secondary hardening and a negative strain rate sensitivity by the maximum stress amplitude and micro-Vickers hardness. During the low cycle fatigue test at the dynamic strain aging regime in inert atmosphere, the variation of cyclic stress amplitude was nearly the same in air except for the occurrence of secondary hardening at the faster strain rate and the higher maximum strength. The earlier appearance of dynamic strain aging in inert atmosphere than in air may be caused by the reduced oxidation effect. Therefore, dynamic strain aging may be suppressed at the initial stage of deformation in air atmosphere. Dynamic strain aging in the oxidizing atmosphere enhanced the number of crack initiation sites by partitioning the local deformation but retarded the crack propagation rate by crack branching and by the suppressed plastic zone size. As the strain rate increased, the fatigue resistance increased at all temperatures. And the effect of dynamic strain aging on fatigue life in a smooth specimen was larger than oxidation effect, while the hardening by

  19. Netherlands Interuniversity Reactor Institut

    International Nuclear Information System (INIS)

    This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

  20. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  1. Management of research reactor ageing

    International Nuclear Information System (INIS)

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

  2. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  3. Análisis de la transmutación de Actínidos Minoritarios en un reactor rápido de sodio con modelo de carga homogéneo mediante el código MCNPX-CINDER

    OpenAIRE

    Ochoa Valero, Raquel; García Herranz, Nuria; Aragonés Beltrán, José María

    2010-01-01

    El reactor rápido refrigerado por sodio (SFR) constituye uno de los conceptos más prometedores de los seis considerados en la Generación IV de reactores nucleares, encontrándose actualmente en fase de investigación. En este marco surge el proyecto europeo CP ESFR (Collaborative Project for an European Sodium Fast Reactor) cuya finalidad es analizar los diversos desafíos y oportunidades que el desarrollo de este tipo de reactores plantea, ya sea en términos de seguridad, tecnología de sodio, c...

  4. China experimental fast reactor

    International Nuclear Information System (INIS)

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  5. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  6. Reactor utilization

    International Nuclear Information System (INIS)

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  7. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  8. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2015-01-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW.

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  11. LWR reactors with supercritical steam states

    International Nuclear Information System (INIS)

    Nuclear reactors cooled with supercritical water are one of the six nuclear systems developed in the context of the Generation IV International Forum. Compared to conventional BWR and PWR reactors, they have higher live steam temperatures above 500 degC, which results a higher specific turbine capacity and a higher performance. The pressure inside the reactor is above the critical pressure of water, so that a boiling crisis in the reactor core is prevented. As in BWR reactors, the superheated steam can be passed on directly into the turbine. It is expected that the power generation cost will be lower than with conventional LWR reactors. One of the challenges is the material for the fuel rod cladding. At peak temperatures above 620 degC, Zircaloy must be replaced by noble steel or ODS. Further, the fuel element boxes must be thermally insulated. Components outside the reactor itself can be the same as in fossil-fuel power stations. (orig.)

  12. Fourth generation type reactors - Synthesis note

    International Nuclear Information System (INIS)

    Six types of reactors have been studied: High or very high temperature helium cooled type reactors, fast neutrons sodium cooled type reactors, fast neutrons gas cooled type reactors, fast neutrons lead or lead-bismuth cooled type reactors, supercritical water type reactors, molten salt type reactors. For the high or very high temperature type reactors the questions of safety and radiation protection have been tackled through the fuel, the neutronics, the materials, the passive systems, safety and reliability of associated industrial processes, risks in relation with graphite, fire and explosion risks linked to hydrogen production; about the fast neutron sodium cooled type reactors the principal questions of safety are tackled through the specific risks linked to the metallic fuel, the neutronic effects in case of loss of coolant said sodium 'vacuum effect', risk of core meltdown, risks linked to sodium, passive systems, ability of structures inspection; concerning the fast neutron gas cooled type reactors, the questions of safety and radiation protection are the aspects linked to the reactor and the aspects linked to the fuel fabrication, this last question has been tackled for each reactor type. A part has been devoted to the production and the management of waste in the case of deployment of a fourth generation reactors park. (N.C.)

  13. Maintenance and material aspects of DREAM reactor

    International Nuclear Information System (INIS)

    A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  15. Precipitation of hydroxyapatite at 37 °C in a meso oscillatory flow reactor operated in batch at constant power density

    OpenAIRE

    Castro, Filipa; Ferreira, António; Rocha, F.A.; A.A. Vicente; Teixeira, J. A.

    2013-01-01

    A meso oscillatory flow reactor (OFR) was successfully applied for the precipitation of hydroxyapatite (HAp) nanoparticles. Mixing efficiency of the mesoreactor operated batchwise in a vertical tube was evaluated at constant power density, by monitoring variation of hue values upon mixing both spatially and temporally. The best operating conditions for fast mixing and a more homogeneous reaction medium were verified for f = 0.83 Hz and x0 = 4.5 mm. HAp precipitation was then carried out under...

  16. Oak ridge test matrix no. 5B and 5C HFR and HFIR irradiations and post-irradiation tensile tests in support of fusion reactor first wall material development

    International Nuclear Information System (INIS)

    In the frame of an ''IEA Implementing Agreement'' between Canada, the EC, Japan, the US and Switzerland, a research and development programme on radiation damage on fusion reactor materials is performed. Annex II of the above agreement defines a series of joint (EC-US-Japan) fusion material irradiations in various MTR's. The present report describes the irradiation device, the irradiation history, the post-irradiation tensile tests, together with comments on the results of these PIE tests, of two of the above irradiations, corresponding to experiments 5B and 5C of the ''Oak Ridge Matrix''. The alloys included were the following: - European Reference AISI316L (''EC316'') - US reference AISI 316L (''US316'') - US P.C.A. (''US-PCA'') - Japanese P.C.A (J-PCA)

  17. The AP1000 reactor

    International Nuclear Information System (INIS)

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  18. Aprovechamiento del permeato de lactosuero para la obtención de ácido L láctico en un reactor de células inmovilizadas.

    OpenAIRE

    Bolumar, Tomás; Monedero, Vicente; Pérez Martínez, Gaspar

    2005-01-01

    El principal objetivo de este trabajo fue establecer un procedimiento piloto para la producción de ácido L-láctico en continuo a partir de permeato de suero lácteo, con el fin de lograr un posterior escalado industrial. Para ello se utilizó un bioreactor con células inmovilizadas de una cepa especialmente seleccionada de Lactobacillus rhamnosus. El sistema es similar al descrito anteriormente (Bruno-Bárcena et al, 1999) y consta de dos reactores, ambos necesarios para su funcionamiento en con...

  19. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  20. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  2. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  5. Performance-based improvement of the leakage rate test program for the reactor containment of HTTR. Adoption of revised test programs containing 'Type A, Type B and Type C tests'

    International Nuclear Information System (INIS)

    The reactor containment of HTTR is periodically tested to confirm leak-tight integrity by conducting overall integrated leakage rate tests, so-called 'Type A tests,' in accordance with a standard testing method provided in Japan Electric Association Code (JEAC) 4203. 'Type A test' is identified as a basic one for measuring whole leakage rates for reactor containments, it takes, however, much of cost and time of preparation, implementation and restoration of itself. Therefore, in order to upgrade the maintenance technology of HTTR, the containment leakage rate test program for HTTR was revised by adopting efficient and economical alternatives including Type B and Type C tests' which intend to measure leakage rates for containment penetrations and isolation valves, respectively. In JEAC4203-2004, following requirements are specified for adopting an alternative program: upward trend of the overall integrated leakage rate due to aging affection should not be recognized; performance criterion for combined leakage rate, that is a summation of local leakage rates evaluated by Type B and Type C tests and converted to whole leakage rates, should be established; the criterion of the combined leakage rate should be satisfied as well as of the overall integrated leakage rate; correlation between the overall integrated and combined leakage rates should be recognized. Considering the historical performances, policies of conforming to the forgoing requirements and of carrying out the revised test program were developed, which were accepted by the regulatory agency. This report presents an outline of the leakage rate tests for the reactor containment of HTTR, identifies practical issues of conventional Type A tests, and describes the conforming and implementing policies mentioned above. (author)

  6. Simulación en código Matlab de un reactor para la gasificación de biomasa en lecho fluidizado burbujeante

    OpenAIRE

    Tosina Fernández, Sergio

    2013-01-01

    Este proyecto simula un reactor unidimensional para la gasificación de biomasa supuesto en régimen estacionario, isotérmico y adiabático, con un esquema simplificado de reacciones, en el que se recogen la teoría hidrodinámica de las “Dos Fases” y distintas correlaciones y parámetros que proceden de ensayos y simulaciones de los últimos cincuenta años, aproximadamente. En los primeros capítulos del proyecto se sientan las bases y conceptos relacionados con la biomasa para una...

  7. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  8. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

  9. Development in UK commercial fast reactor design

    International Nuclear Information System (INIS)

    The design of the CDFR commercial demonstration fast reactor which should be put into operation early in the 90-ties is described. Basic elements of the reactor components are considered. The choice of the integrated primary coolant circuit, and design of intermediate heat exchangers, sodium pumps and charging machines is substantiated. The reactor power is 1320 MW(e), or 3300 MW(t). The sodium temperature at the reactor inlet is 370 deg C, at its outlet 540 deg C. Linear loading per fuel element length is 40 W/mm. The conclusion is drawn that the described design of the demonstration reactor fully corresonds to requirements of a full-scale commercial NPP with a fast reactor

  10. Decay of reactor neutrinos

    International Nuclear Information System (INIS)

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Goesgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit t/sup c.m.//m/sub ν/ is 1--3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1--4 MeV could decay into electron-positron pairs. These pairs were not observed, and from the absence of such a signal we deduce restrictions on the corresponding mixing parameters

  11. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  12. Packed-Bed Reactor Study of NETL Sample 196c for the Removal of Carbon Dioxide from Simulated Flue Gas Mixture

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, James S.; Hammache, Sonia; Gray, McMahan L.; Fauth Daniel J.; Pennline, Henry W.

    2012-04-24

    An amine-based solid sorbent process to remove CO2 from flue gas has been investigated. The sorbent consists of polyethylenimine (PEI) immobilized onto silica (SiO2) support. Experiments were conducted in a packed-bed reactor and exit gas composition was monitored using mass spectrometry. The effects of feed gas composition (CO2 and H2O), temperature, and simulated steam regeneration were examined for both the silica support as well as the PEI-based sorbent. The artifact of the empty reactor was also quantified. Sorbent CO2 capacity loading was compared to thermogravimetric (TGA) results to further characterize adsorption isotherms and better define CO2 working capacity. Sorbent stability was monitored by periodically repeating baseline conditions throughout the parametric testing and replacing with fresh sorbent as needed. The concept of the Basic Immobilized Amine Sorbent (BIAS) Process using this sorbent within a system where sorbent continuously flows between the absorber and regenerator was introduced. The basic tenet is to manipulate or control the level of moisture on the sorbent as it travels around the sorbent circulation path between absorption and regeneration stages to minimize its effect on regeneration heat duty.

  13. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  15. Characteristics of radiation porosity formed upon irradiation in a BN-600 reactor in the fuel-element cans of cold-deformed steel EK-164 (06Kh16N20M2G2BTFR)-ID c.d.

    Science.gov (United States)

    Portnykh, I. A.; Kozlov, A. V.; Panchenko, V. L.; Mitrofanova, N. M.

    2012-05-01

    At present, it is the austenitic cold-deformed steel EK164 (06Kh16N20M2G2BTFR)-ID that is considered as a promising material for the achievement of a maximum damage (no less than 110 dpa) and maximum burnup (≥15%). In this work, we have determined the characteristics of porosity formed upon irradiation in a BN-600 reactor to the maximum damaging dose of 77 dpa in the materials of fuel-element cans made of cold-deformed steel EK164-ID c.d. A comparison has been made with analogous characteristics obtained earlier using the standard material, i.e., the cold-deformed steel ChS68 (06Kh16N 15M2G2TFR)-ID c.d.

  16. Nuclear research reactors

    International Nuclear Information System (INIS)

    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.)

  17. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  19. Mass and energy balance: application to the sanitary sewage treatment with an upflow anaerobic sludge blanket (UASB) reactors to temperature of 20 deg C; Balanco de massa e energia: aplicacao ao tratamento de esgotos sanitarios com reatores anaerobicos de manta de lodo (UASB) a temperatura de 20 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Carrasco, Katherine Ivonne Alcocer

    1992-07-01

    A feasibility study of an Upflow Anaerobic Sludge Blanket (UASB) reactor use as a sewage treatment plant component for areas with average temperature bellow 20 deg C was performed. The literature on UASB reactor indicates that a 70 % chemical oxygen demand (COD) removal al 20 and 6 to 17 hr hydraulic detention time is possible in sewage treatment plants. This study was developed for a Oruro (Bolivia) district and the plant was designed for a population of up to 10,000 inhabitants. This city presents average temperatures lower than 20 deg C being necessary to heat the sewage if is used the UASB reactor. Based on the performance simulation of mass and energy balances it was found that 84 % COD removal and 92 % total suspended solids removal are possible. The potential average energy production (61 kW due to methane combustion) is less than 10 % of the power consumption for heating, which indicates that the use of the methane may be expensive. The evaluated energy rate to be applied to the sewage for heating is 0.33 kW/m{sup 3} d{sup -1} which is significantly greater than the necessary energy to introduce oxygen in aerobic treatment systems. However total energy demand for aerobic systems must be evaluated for each particular case. (author)

  20. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  1. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  2. Results of charpy V-notch impact testing of structural steel specimens irradiated at ∼30 degrees C to 1 x 1016 neutrons/cm2 in a commercial reactor cavity

    International Nuclear Information System (INIS)

    A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at ∼ 30 degrees C (∼ 85 degrees F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 1016 neutrons/cm2 (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was ∼ 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of ∼ 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications

  3. Fast reactors: the future of nuclear energy

    International Nuclear Information System (INIS)

    The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.)

  4. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  5. Corrosion of graphitic high temperature reactor materials in steam/helium mixtures at total pessures of 3-55 bar and temperatures of 900-1150 C (1173-1423K)

    International Nuclear Information System (INIS)

    In course of accident examination for (HTR), experiments on the corrosion behavior of graphitic reactor materials in steam have been performed a total pressures of 3-55bar and temperatures of 900-1150 C (1173-1423K); these experiments and their evaluation are documented here. Reactor materials examined are the structure graphite V483T2 and the fuel element matrices A3-27 and A3-3. In all experiments, the steam partial pressure was 474mbar (inert gas helium). The dependence of reaction rates and density profiles on burn-off, total pressure and temperature has been examined. Experimental reaction rates depending on burn-off are fitted by theoretical curves, a procedure, which allows rate comparison for a well defined burn-off. Comparing rates as a function of total pressure, V483T2 shows a linear dependence on 1√ptotal, whereas for matrix materials a pressure independent rate was found for ptotal4mm for A3-3. (orig.)

  6. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  7. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  8. Watercooled atomic reactor

    International Nuclear Information System (INIS)

    A hydraulic method for holding fuel assemblies down is described. The nuclear reactor has the vessel divided into an upper high pressure plenum and a lower low pressure or outlet plenum by means of a seal plate. A piston is located at the seal plate with its upper surface exposed to the high pressure and its lower surface exposed to the lower pressure. A push rod attached to the piston extends downwardly in butting relationship with fuel assemblies to hold them down. (C.F.)

  9. Skin Inqjuries Reduce Survival and Modulate Corticosterone, C-Reactive Protein, Complement Component 3, IgM, and Prostaglandin E2 after Whole-Body Reactor-Produced Mixed Field (n + γ-Photons Irradiation

    Directory of Open Access Journals (Sweden)

    Juliann G. Kiang

    2013-01-01

    Full Text Available Skin injuries such as wounds or burns following whole-body γ-irradiation (radiation combined injury (RCI increase mortality more than whole-body γ-irradiation alone. Wound-induced decreases in survival after irradiation are triggered by sustained activation of inducible nitric oxide synthase pathways, persistent alteration of cytokine homeostasis, and increased susceptibility to systemic bacterial infection. Among these factors, radiation-induced increases in interleukin-6 (IL-6 concentrations in serum were amplified by skin wound trauma. Herein, the IL-6-induced stress proteins including C-reactive protein (CRP, complement 3 (C3, immunoglobulin M (IgM, and prostaglandin E2 (PGE2 were evaluated after skin injuries given following a mixed radiation environment that might be found after a nuclear incident. In this report, mice received 3 Gy of reactor-produced mixed field (n+γ-photons radiations at 0.38 Gy/min followed by nonlethal skin wounding or burning. Both wounds and burns reduced survival and increased CRP, C3, and PGE2 in serum after radiation. Decreased IgM production along with an early rise in corticosterone followed by a subsequent decrease was noted for each RCI situation. These results suggest that RCI-induced alterations of corticosterone, CRP, C3, IgM, and PGE2 cause homeostatic imbalance and may contribute to reduced survival. Agents inhibiting these responses may prove to be therapeutic for RCI and improve related survival.

  10. Welcome address to 30th international meeting on Reduced Enrichment for Research and Test Reactors, 6 October 2008, Washington, D.C., USA

    International Nuclear Information System (INIS)

    The RERTR meetings, of course, have a special role among our activities. They are important not only for what is presented in the formal technical sessions, but also for what takes place in the informal gatherings, side meetings, and personal encounters. The IAEA also has participated as an observer in the International Fuel Development Working Group, a highly important effort that is overseeing the multinational cooperative research and development effort for very high density LEU fuel. Remarkable achievements of GTRI are the conversion to LEU or final shutdown prior to conversion of 62 research reactors, the return to the United States of more than 1100 kg of spent HEU fuel and more than 1800 kg of spent LEU fuel and the return to Russia of more than 600 kg of spent and fresh HEU fuel. But while much has been achieved so far, vulnerabilities remain. HEU continues to be used for military purposes in a number of States; about 150 civilian and military research reactors are still using HEU and important quantities of fresh; and spent HEU fuel continues to be stored in different countries. All this calls for continued efforts, with a sense of urgency and more coherent global action. Some of the measures that might be taken are as follows: The countries involved should join forces to step up their efforts towards minimizing and eventually eliminating the civilian and in due course the military use of HEU. Financing and other incentives should be made available where needed to assist countries with conversion operations. All countries should agree to stop producing fissile material for use in nuclear weapons. The elements are already in place for such an agreement, in the form of the proposed Fissile Material Cut-off Treaty. It is high time to negotiate and conclude such a treaty. To build confidence, countries with civilian and military HEU stockpiles should declare the size of those stockpiles and publish a schedule under which the remaining HEU will be verifiably

  11. Political-social reactor problems at Berkeley

    International Nuclear Information System (INIS)

    For better than ten years there was little public notice of the TRIGA reactor at UC-Berkeley. Then: a) A non-student persuaded the Student and Senate to pass a resolution to request Campus Administration to stop operation of the reactor and remove it from campus. b) Presence of the reactor became a campaign-issue in a City Mayoral election. c) Two local residents reported adverse physical reactions before, during, and after a routine tour of the reactor facility. d) The Berkeley City Council began a study of problems associated with radioactive material within the city. e) Friends Of The Earth formally petitioned the NRC to terminate the reactor's license. Campus personnel have expended many man-hours and many pounds of paper in responding to these happenings. Some of the details are of interest, and may be of use to other reactor facilities. (author)

  12. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  13. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  14. Photo-removal of sulfamethoxazole (SMX) by photolytic and photocatalytic processes in a batch reactor under UV-C radiation ({lambda}{sub max} = 254 nm)

    Energy Technology Data Exchange (ETDEWEB)

    Nasuhoglu, Deniz; Yargeau, Viviane [Department of Chemical Engineering, McGill University, 3610 University Street, Montreal, Quebec, H3A 2B2 (Canada); Berk, Dimitrios, E-mail: dimitrios.berk@mcgill.ca [Department of Chemical Engineering, McGill University, 3610 University Street, Montreal, Quebec, H3A 2B2 (Canada)

    2011-02-15

    In this study, photolytic and photocatalytic removal of the antibiotic sulfamethoxazole (SMX) under UVC radiation ({lambda} = 254 nm) was investigated. The light intensity distribution inside the batch photoreactor was characterized by azoxybenzene actinometry. The intensity of incident radiation was found to be a strong function of position inside the reactor. 12 mg L{sup -1} of SMX was completely removed within 10 min of irradiation under UVC photolysis, compared to 30 min under TiO{sub 2} photocatalysis. COD measurement was used as an indication of the mineralization efficiency of both processes and higher COD removal with photocatalysis was shown. After 6 h of reaction with photolysis and photocatalysis, 24% and 87% removal of COD was observed, respectively. Two of the intermediate photo-products were identified as sulfanilic acid and 3-amino-5-methylisoxazole by direct comparison of the HPLC chromatograms of standards to those of treated solutions. Ecotoxicity of treated and untreated solutions of SMX towards Daphnia magna was also investigated. It was found that a 3:1 ratio of sample to standard freshwater and a high initial concentration of 60 mg L{sup -1} of SMX were used to obtain reliable and reproducible results. The photo-products formed during photocatalytic and photolytic processes were shown to be generally more toxic than the parent compound.

  15. Photo-removal of sulfamethoxazole (SMX) by photolytic and photocatalytic processes in a batch reactor under UV-C radiation (λmax = 254 nm)

    International Nuclear Information System (INIS)

    In this study, photolytic and photocatalytic removal of the antibiotic sulfamethoxazole (SMX) under UVC radiation (λ = 254 nm) was investigated. The light intensity distribution inside the batch photoreactor was characterized by azoxybenzene actinometry. The intensity of incident radiation was found to be a strong function of position inside the reactor. 12 mg L-1 of SMX was completely removed within 10 min of irradiation under UVC photolysis, compared to 30 min under TiO2 photocatalysis. COD measurement was used as an indication of the mineralization efficiency of both processes and higher COD removal with photocatalysis was shown. After 6 h of reaction with photolysis and photocatalysis, 24% and 87% removal of COD was observed, respectively. Two of the intermediate photo-products were identified as sulfanilic acid and 3-amino-5-methylisoxazole by direct comparison of the HPLC chromatograms of standards to those of treated solutions. Ecotoxicity of treated and untreated solutions of SMX towards Daphnia magna was also investigated. It was found that a 3:1 ratio of sample to standard freshwater and a high initial concentration of 60 mg L-1 of SMX were used to obtain reliable and reproducible results. The photo-products formed during photocatalytic and photolytic processes were shown to be generally more toxic than the parent compound.

  16. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  17. REUSABLE ADSORBENTS FOR DILUTE SOLUTIONS SEPARATION. 6. BATCH AND CONTINUOUS REACTORS FOR ADSORPTION AND DEGRADATION OF 1,2-DICHLOROBENZENE FROM DILUTE WASTEWATER STREAMS USING TITANIA AS A PHOTOCATALYST. (R828598C753)

    Science.gov (United States)

    Two types of external lamp reactors were investigated for the titania catalyzed photodegradation of 1,2-dichlorobenzene (DCB) from a dilute water stream. The first one was a batch mixed slurry reactor and the second one was a semi-batch reactor with continuous feed recycle wit...

  18. Five Lectures on Nuclear Reactors Presented at Cal Tech

    Science.gov (United States)

    Weinberg, Alvin M.

    1956-02-10

    The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

  19. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  20. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  1. RB reactor noise analysis

    International Nuclear Information System (INIS)

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  2. 75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Science.gov (United States)

    2010-12-20

    ... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR... Director, Office of Nuclear Reactor Regulation under 10 CFR 50.4. In addition, licensee submittals...

  3. The relationship between dose rate and transformation induction in C3H/10T1/2 cells by TRIGA reactor fission neutrons at 0.3 Gy

    International Nuclear Information System (INIS)

    The authors present their own and other data showing dose-effect relations for cell survival and the induction of transformations in C3H/IOT 1/2 cells in exponential or stationary cultures after a range of high dose-rate irradiations with X-rays or AFRRI neutrons. (UK)

  4. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  5. Nordic study on reactor waste

    International Nuclear Information System (INIS)

    In 1981, 14 nuclear power reactors are in operation and 2 under construction in the Nordic countries. So far, the reactor waste originating from day-to-day operation of these plants has been stored in solidified form at the reactor sites. Within a few years a satisfactory disposal procedure needs to be established. While the main R and D effects in the waste field have earlier been devoted to the question of irradiated fuel and waste from reprocessing, there is therefore now an increased interest in reactor waste with its much lower radioactivity but somewhat larger volumes. Since 1977, efforts have been made in a joint Nordic study to examine which facts need to be known in order to perform a comprehensive safety assessment of a reactor waste management system. In the present study a Reference system related to the waste generated over 30 years from six 500 MW-reactors is examined. The dominating radionuclides during storage and transportation accident scenarios are Cs-134, Cs-137 and Co-60. For most of the release scenarios from repositories Cs-137 and Sr-90 are dominating. Some scenarios are, however, dominated by the very longlived nuclides I-129 and C-14. A closer examination of the concentration in the waste of these nuclides and of their leaching properties indicates that their small - but significant - influence, as calculated, is probably grossly overestimated. The mechanical stability obtained in routine solidification processes of reactor waste products in conjunction with the outer container (steel drum, transport container, etc.) turns out to be sufficient. Difficulties were encountered in applying ICRP methodology and available dose calculation methods to calculation of population doses due to small activity releases, and effects extending into the far future. (EG)

  6. Reactor Physics Training

    International Nuclear Information System (INIS)

    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

  7. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  8. Analisis de un transitorio de inyección de Boro en un reactor PWR con el código acoplado RELAP5/PARCSv2.7

    OpenAIRE

    Garcia-Fenoll, Marina; Abarca Giménez, Agustín; Barrachina Celda, Teresa María; Miró Herrero, Rafael; Verdú Martín, Gumersindo Jesús

    2011-01-01

    En este trabajo se presenta la implementación de una nueva prestación en el código acoplado RELAP5/PARCS v2.7 que permite analizar transitorios en los que se produce una variación de a concentración de boro en el núcleo. La implementación de la opción de inyección/dilución de boro consiste en la modificación del código fuente para que sea capaz de utilizar tablas de secciones eficaces distintas para diversas concentraciones de boro e interpolar entre ellas, así como en la mejora de la informa...

  9. Safeguarding research reactors

    International Nuclear Information System (INIS)

    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

  10. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  11. Electroceramic reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bagger, C. [Risoe National Lab., Materials Research Dept. (Denmark)

    1999-10-01

    Production of Gd-doped and Y-doped ceria has been successfully accomplished using a continuous technique with industrial production potential. Production parameters for tape casting and low temperature sintering of Gd-doped ceria membranes have been established as well. Yttria doping has been found cheaper than gadolinia doping, but sintering to gastightness was difficult, because grain growth is suppressed. The volume stability at 600 deg. C of yttria doped ceria during reduction was high. (EHS)

  12. Effects of type of reactor, crystallinity of SiC, and NF3 gas pressure on etching rate and smoothness of SiC surface using NF3 gas plasma

    International Nuclear Information System (INIS)

    Polycrystalline β-SiC and single-crystalline 4H-SiC surfaces were etched by reactive ion etching (RIE) using NF3 gas plasma. A smooth surface was obtained on the polycrystalline SiC after RIE at NF3 gas pressures of 2 and 10 Pa for 10 min, and neither spikes nor pillars were formed on it. On the other hand, some pillars were formed on the single-crystalline SiC surface by RIE at NF3 gas pressures of 2 and 10 Pa. Though the absence of carbon-rich regions and SiOx on the outermost surface before etching was confirmed by x-ray photoelectron spectroscopy and Raman analysis, x-ray diffraction analysis revealed that graphite crystallites were present in the single-crystalline SiC bulk. It was concluded that the graphite crystallites acted as masks and the pillars grew up from the graphite crystallites in the single crystalline SiC during RIE.

  13. Uptake of [35S] carbonyl sulphide and [14C] carbon dioxide by crops in the vicinity of an advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    The uptake into major crops of 35S and 14C released to the atmosphere during operation of Hinkley Point Nuclear Power Station was studied at an on-site experimental plot. Both radionuclides were measured in air and edible crop parts, while the effect of boiling on the 35S content of the latter was also investigated. The results were analysed in terms of air to crop transfer factors and the implications of these were assessed for both collective dose and dose to a hypothetical critical group. The transfer factor for 35S to green vegetables was found to be much smaller than believed previously, to the extent that the vegetable consumption pathway is of secondary importance to that of fresh milk. A possible reduction is indicated in critical group individual 35S doses of 20-60% and in collective dose by a factor of about 2, but vegetable consumption remains of potential radiological significance. It was confirmed that the specific activity approach used currently for assessments of 'first pass' dose from 14C releases is broadly correct, but a possible reduction of up to 3.5 times is indicated in the air-crop transfer for root vegetables. (author)

  14. Mechanical properties of a high-purity Fe-9Cr-2W-0.1C model alloy for low-activation ferritic steels for fusion reactors

    International Nuclear Information System (INIS)

    Reduced-activation ferritic steels, such as JLF-1 (Fe-9Cr-2W-0.1C-V-Ta-N) steel, are candidate structural materials for use in the next fusion device. It is essential to develop low-activation ferritic steels which have been purified in order to reduce the amount of radioactive impurity elements. It is important to clarify whether or not purification causes any significant changes in the mechanical properties of the ferritic steels, as they will be used as structural materials. In this study a high-purity Fe-9Cr-2W-0.1C model alloy was prepared from an ultra-high-purity iron. A significantly lower tensile strength was observed in the high-purity ferritic steel at room temperature and 573 K, as compared to that of JLF-1 steel. The difference in yield strength at room temperature is considered to be mainly due to a lower content of nitrogen and soluble impurities, a larger packet diameter and a larger lath length in the case of the high-purity steel. The hardness of the as-quenched high-purity steel and the recovery behavior below 823 K are not significantly different from those of conventional ferritic steel. (author)

  15. Proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic

    International Nuclear Information System (INIS)

    The proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic - 6. ENFIR - allow to evaluate the present status of development in reactor physics and thermohydraulic fields. The mathematical models and methods for calculating neutronic of nuclear reactors, safety reactor analysis, measuring methods of neutronic parameters, computerized simulation of accidents, transients and thermohydraulic analysis are presented. (M.C.K.)

  16. SP-100 Reactor Subsystem Development

    Science.gov (United States)

    Demuth, Scott F.

    1994-07-01

    The SP-100 reactor subsystem consists of the pressure vessel, vessel internals, and fuel elements. Type A (standard) Nb-1Zr and rhenium materials development efforts related to fabrication of the vessel, vessel internals, and fuel cladding/liner have been completed. Type A and Type C (PWC-11) Nb-1Zr loop fabrication has been successfully demonstrated by prototypic testing with flowing lithium at 1350 K for 1500 hr. Development of UN fuel has been completed, and the performance validated by irradiation testing to the full life (7 yr. full power) burnup of 6 atom %. Neutronic and hydraulic core performance have been validated by engineering mockup critical experiments in the Zero Power Physics Reactor at Argonne National Laboratory, and detailed core hydraulic flow testing with water. Essentially all feasibility issues have been settled for the full life SP-100 reactor subsystem. Remaining SP-100 reactor subsystem development efforts are focused on further reducing mass by the use of Type C (PWC-11) Nb-1Zr rather than Type A, and demonstrating fuel life for beyond full life to perhaps 9 atom % burnup.

  17. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO2. The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  18. Reactor Physics Programme

    International Nuclear Information System (INIS)

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  19. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  20. Evaluation of corrosion of 800GN alloy tubes in similar ambient to the secondary circuit of PWR reactor at 80 deg C

    International Nuclear Information System (INIS)

    In this work we investigated the effect of the presence of chloride ions (concentrations of 10, 50 and 250 ppb) and sulfate in a ratio of 1: 1 in a corrosion behavior of 800NG alloy tube to 80 ° C, in electrochemical cell to three electrodes. Such concentrations correspond to action levels used in the pipeline safety. Experiments in potential open circuit and cyclic polarization were used to characterize the corrosion behavior of the material. Morphological analysis of corrosion and corrosion products was performed by optical microscopy, scanning electron microscopy and X-ray diffraction. The results showed that the morphology of the attack located in a alloy 800GN is related to the ratio between the concentrations of chloride and sulfate ions in the medium

  1. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  2. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  3. Reactor System Design

    International Nuclear Information System (INIS)

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  4. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  5. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  6. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    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)

  7. Solid-state Fermentation of Xylanase from Penicillium canescens 10-10c in a Multi-layer-packed Bed Reactor

    Science.gov (United States)

    Assamoi, Antoine A.; Destain, Jacqueline; Delvigne, Frank; Lognay, Georges; Thonart, Philippe

    Xylanase is produced by Penicillium canescens 10-10c from soya oil cake in static conditions using solid-state fermentation. The impact of several parameters such as the nature and the size of inoculum, bed-loading, and aeration is evaluated during the fermentation process. Mycelial inoculum gives more production than conidial inoculum. Increasing the quantity of inoculum enhances slightly xylanase production. Forced aeration induces more sporulation of strain and reduces xylanase production. However, forced moistened air improves the production compared to production obtained with forced dry air. In addition, increasing bed-loading reduces the specific xylanase production likely due to the incapacity of the Penicillium strain to grow deeply in the fermented soya oil cake mass. Thus, the best cultivation conditions involve mycelial inoculum form, a bed loading of 1-cm height and passive aeration. The maximum xylanase activity is obtained after 7 days of fermentation and attains 10,200 U/g of soya oil cake. These levels are higher than those presented in the literature and, therefore, show all the potentialities of this stock and this technique for the production of xylanase.

  8. High solids fermentation reactor

    Science.gov (United States)

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  9. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  10. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  11. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  12. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  13. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  14. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  15. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  16. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  17. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  18. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  19. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  20. Requirements for light water reactors

    International Nuclear Information System (INIS)

    The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10-6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

  1. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  2. Investigation of Ultrasound Assisted Nucleation in a Milliflow Reactor

    OpenAIRE

    Jordens, Jeroen; Janssen, Geert; Van Gerven, Tom; Braeken, Leen

    2013-01-01

    The purpose of our research was to perform the cooling crystallization of paracetamol in a milliflow reactor and study the effect of the ultrasound frequency and power on the metastable zone width (MSZW) and the crystal size distribution (CSD). Without ultrasound, the MSZW was reduced by 11.5 °C when using the milliflow reactor instead of the batch reactor. Applying ultrasound resulted for both reactor setups in a significant reduction of the MSZW. The smallest MSZW (2.9°C) was observed in...

  3. Study of the hydrogen behavior in amorphous hydrogenated materials of type a - C:H and a - SiC:H facing fusion reactor plasma; Etude du comportament de l`hydrogene dans des materiaux amorphes hydrogenes de type a - C:H et a - SiC:H devant faire face au plasma des reacteurs a fusion

    Energy Technology Data Exchange (ETDEWEB)

    Barbier, G. [Lyon-1 Univ., 69 - Villeurbanne (France). Inst. de Physique Nucleaire

    1997-04-10

    Plasma facing components of controlled fusion test devices (tokamaks) are submitted to several constraints (irradiation, high temperatures). The erosion (physical sputtering and chemical erosion) and the hydrogen recycling (retention and desorption) of these materials influence many plasma parameters and thus affect drastically the tokamak running. First, we will describe the different plasma-material interactions. It will be pointed out, how erosion and hydrogen recycling are strongly related to both chemical and physical properties of the material. In order to reduce these interactions, we have selected two amorphous hydrogenated materials (a-C:H and a-SiC:H), which are known for their good thermal and chemical qualities. Some samples have been then implanted with lithium ions at different fluences. Our materials have been then irradiated with deuterium ions at low energy. From our results, it is shown that both the lithium implantation and the use of an a - SiC:H substrate can be beneficial in enhancing the hydrogen retention. These results were completed with thermal desorption studies of these materials. It was evidenced that the hydrogen fixation was more efficient in a-SiC:H than in a-C:H substrate. Results in good agreement with those described above have been obtained by exposing a - C:H and a - SiC:H samples to the scrape off layer of the tokamak of Varennes (TdeV, Canada). A modelling of hydrogen diffusion under irradiation has been also proposed. (author) 176 refs.

  4. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  5. Thai research reactor

    International Nuclear Information System (INIS)

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  6. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  7. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    fuels originates from goals for achieving high burnup, operating at higher temperature, and the incorporation of the minor actinides (Np, Am, Cm) into the fuels. High burn-ups will allow uninterrupted reactor operations over longer periods of time and consequently, reduction of spent fuel volumes, and eventually a significant fuel cycle reduction cost. High burn-ups are however associated with physical limitations which are primary due to the swelling of the fuel and oxidation of cladding inner surface as well as the dimensional stability of core materials such as cladding and subassembly duct due to high fast neutron dose. Higher temperature operation also challenges the performance of cladding materials and hence advanced cladding materials are needed for high temperature operation. The irradiation performance database for (U,Pu)N mixed nitride (MN) fuels is substantially smaller than that for metal carbide (MC) fuels, and these fuels can be considered to be at an early stage of development relative to oxide and metal fuels. Compared to MC fuels, MN fuels exhibit less fuel swelling, lower fission gas release, however, the problem of the production of biologically hazardous 14C in nitride fuels fabricated using natural nitrogen poses a considerable concern for the nitride spent fuel waste management. Interest remains in nitride fuels due to the combination of high thermal conductivity and high melting point. The paper also addresses the technology readiness level (TRL) concept as applied to various fuel options. (author)

  8. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  9. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  10. Power calibration study at the Musashi reactor

    International Nuclear Information System (INIS)

    The Musashi reactor (TRIGA-II,100 kW) initially went critical in January of 1963. The reactor had been used for training, isotope production and medical irradiation for boron neutron capture therapy (1). The initial power calibration was based on the use of a calibrated electrical heater in a calorimetric procedure where the rate of rise of the bulk pit water temperature was measured using 2 kW heaters x 6 pieces. The rate of rise of water temperature was determined to be 0.0474 C/kWh. The reactor was then operated to give the same rate of rise of water temperature. Thus the reactor power was established at the value produced by the electrical heaters. A stirrer for tank water mixing was not used. Recent communications (2)(3) indicated that power calibrations using a stirrer provided a much more uniform mixing, and heating in the reactor tank water which was essential for an accurate calibration. In this paper, the effect of mixing using a stirrer was investigated considering the physical factors such as room temperature, humidity, tank water temperature and it's distributions. The room temperature and humidity around the reactor varies 6-30 and 30-80 %, respectively, depending on four seasons. The heat flow through the surface of the pool was also evaluated because the reactor usually operates without cover on the surface of the pool. (orig.)

  11. The Swedish Zero Power Reactor R0

    International Nuclear Information System (INIS)

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

  12. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-15

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  13. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  14. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 oC average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  15. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  16. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  17. Multipurpose research reactors

    International Nuclear Information System (INIS)

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  18. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  19. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  20. Single purpose reactor for sea water desalination

    International Nuclear Information System (INIS)

    Some possibilities of a single-purpose reactor for sea water desalination are outlined. Preliminary economic evaluations are also presented and emphasis is given to the prospects of a simplified reactor for sea water desalination. Because no more than 100M3/year are required in one place at one time and given the lack of experience in operating very large desalination plants, it seems that the single-purpose reactor should be small (between 200 mwt to 600 mwt). Two new concepts for desalination plants have been recently developed in Israel: an aluminium horizontal tube multieffect evaporator (AHTME) designed and manufactured by Israel Desalination Engineering Ltd. and a direct contact condensation (DCC) plant, whose 50,000 GPD pilot plant is under construction in the city of Haifa. These two concepts of desalination plants are characterized by economy and operating temperatures below 100 deg C. For the AHTME, the optimum water cost corresponds to a steam temperature of about 70 deg C. A water-cooled reactor can be employed without the need of pressurized vessels or tubes; in addition, cheap construction materials such as aluminium can be used. The advantages of combining a simplified reactor and improved desalination plants and the advantages of the single purpose reactor could bring about a cheaper cost for desalinated water. (author)

  1. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  2. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  3. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  4. PFBR reactor protection

    International Nuclear Information System (INIS)

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  5. Reactor BR2

    International Nuclear Information System (INIS)

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  6. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  7. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  8. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  9. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  10. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  11. Heat pipe reactors for space power applications

    Science.gov (United States)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  12. Thermionic reactor systems for electric propulsion.

    Science.gov (United States)

    Mondt, J. F.

    1972-01-01

    This paper summarizes the preliminary design studies of unmanned electric propulsion spacecraft, with primary emphasis on the in-core thermionic reactor power subsystem. A 70-kWe power subsystem, with an external-fuel thermionic reactor, is shown integrated into a large L/D (about 20) electric propulsion spacecraft. The 70-kWe spacecraft is designed for launch to earth escape with a Titan-Centaur. Two 300-kWe reactor designs (external-fuel and flashlight designs from Atomic Energy Commission contracted studies) are integrated into 270-kWe electric propulsion spacecraft. The 270-kWe spacecraft are designed for launch to a 700-nmi earth orbit with a Titan III-C/7 booster. The 70-kWe thermionic reactor power subsystem is also conceptually shown as a space base power plant.

  13. Design of a nuclear reactor cooperative controller

    International Nuclear Information System (INIS)

    This paper describes the development of a fuzzy logic controller software package and explores the feasibility of its use in nuclear reactor operation. The controller complements reactor operator actions, and the operators can override the controller decisions. Techniques of providing learning capability to the controller are also being investigated to improve the reasoning and control skill of the controller. The fuzzy logic controller is implemented in C language and its overall structure is shown. The heart of the systems consists of a fuzzifier, a rule interpreter, and a defuzzifier. The controller is designed as a stand-alone package that can be interfaced to a simulated model of a nuclear reactor. Since no model is an accurate representation of the actual process being modeled, some tuning must be performed to use the controller in an actual reactor. This is accomplished using the learning feature of the controller

  14. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  15. Reactor construction steels

    International Nuclear Information System (INIS)

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  16. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  17. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  18. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  19. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  20. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)

  1. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  2. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  3. Optimized Design and Discussion on Middle and Large CANDLE Reactors

    OpenAIRE

    Xiaoming Chai; Yong Zhang; Mingyu Yan

    2012-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor) reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scal...

  4. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 5500C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  5. High temperature reactors and their use in the FRG

    International Nuclear Information System (INIS)

    Various aspects of the strategy of building high temperature reactors in the FRG are discussed. The development of these reactors has a long tradition in the FRG and great sums of money are being invested in the research programme. In 1988 the AVR-15 experimental reactor is expected to be shut down in which the helium output temperature had been maintained at 950 degC for a long period of time. The THTR-300 demonstration power plant which is expected to be available at that time represents a link to further application of high temperature reactors in the FRG. A detailed description is presented of projects of further high temperature reactors with a wide range of power output. The BBC/HRB association with Swiss participation is now specifying the project of the HTR-500 reactor with a steam cycle and the delivery of technological steam. This reactor should be followed up by the construction of a reactor with an HHT gas turbine and of an HTR-PNP reactor for coal gasification. Alternatively developed are small HTR-100 universal reactors. Prospective projects also include the 80 MW modular system by KWU following up on the AVR-15 reactor. (Z.M.)

  6. Studies on transferring the safety features of the module reactor to a large power reactor

    International Nuclear Information System (INIS)

    The German industries and research institutions have developed the HTR module reactor, which is strongly characterized by inherent safety features. The power output is limited to about 200 MWth because of its core configuration. It has been investigated in this work, whether the safety features of the module reactor can be transferred to larger power reactors. For this purpose the conceptual design of a ring core pebble bed reactor has been made with a thermal power output of 3000 MW. By means of computer calculations, the principal physical, thermohydraulical and safety features of the ring reactor have been studied. It has been shown that the 3000-MWth ring reactor basically possesses the same safety characteristics as the small module reactor. At reactivity disturbances, the reactor is shut down passively by the strongly negative temperature coefficient. The decay heat removal is also realized based on the passive priniciple. In the case of a total loss of coolant, the maximum fuel element temperature remains below 1600deg C; and consequently the retention of fission products in the fuel elements is fully attained. The control of xenon oscillations takes place inherently due to the mutual coupling between the local power production and the fuel temperature. (orig.)

  7. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    arising in nuclear field and especially in works implying research reactors result first from the synthesis of the problems which sometimes are conventionally treated depending on the experience of the decision staff. Abnormal or un-specific problems from the technical point of view but always with economic consequences, as risk doses may occur. A series of such aspects and corresponding measures are discussed for the different situations as follows: a. Startup, operation, and shutdown of the reactor and, where appropriate, experimental devices; b. Loading, unloading, and movement within the reactor of fuel and other core and reflector components, including experimental devices; c. Routine maintenance of major components or systems that could have an effect on reactor safety; d. Inspections and tests of structures, systems and components that may have an effect on reactor safety, including those specified in the approved programme of periodic testing and inspection; e. Personnel radiation protection consistent with applicable regulations; f. Authorization of operation and maintenance and the conduct of irradiations and experiments that could affect reactor safety or radioactivity; g. Operator response to appropriate anticipated operational occurrences and, to the extent feasible, accident conditions; h. Emergency actions; i. Safety issues. Finally the handling of radioactive wastes and control monitoring of radioactive release are discussed

  8. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    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.)

  9. Development of coated fuel particle for VHTR reactor

    International Nuclear Information System (INIS)

    The VHTR reactor is a high temperature reactor concept with a 1000oC outlet temperature and it is one of IV generation reactor that is being developed. This reactor is developed based on GT-MHR reactor with higher outlet temperature reactor, burn-up and electricity efficiency. Accordingly, material improvement is necessary, such as: SiC in correlation to elevated temperature of 850oC to 1000oC. The SiC layer plays an important role. Beside retaining metallic fission products (Cs-137, Ru-106) released from fuel kernel, it also provides mechanical strength and dimension stability to fuel particle. The ZrC layer can be considered to replace SiC because of its higher temperature capability. ZrC has a melting point of 35400C whereas SiC has a melting point of 1800oC. The ZrC-TRISO coated fuel particle did not fail until ∼6000 seconds at 2200oC heating, while a few percent of the conventional TRISO-coated fuel particles failed already by 2200oC, and almost 100% instantaneously at 2400oC. In the whole temperature range, ZrC layer showed more resistance to chemical corrosion by fission products, especially by palladium (Pd) which is known to corrode the SiC layer. However, The ZrC layer capability to retention of Ru-106 was inferior to SiC. The retention of the other metal fission products such as barium, silver and promethium appears to be better than SiC. (author)

  10. One piece reactor removal

    International Nuclear Information System (INIS)

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  11. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  12. The fusion reactor

    International Nuclear Information System (INIS)

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  13. Polymerization Reactor Engineering.

    Science.gov (United States)

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  14. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  15. Gas-cooled reactors

    International Nuclear Information System (INIS)

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  16. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  17. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  18. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  19. Naval propulsion reactors

    International Nuclear Information System (INIS)

    This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)

  20. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  1. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  2. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  3. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  4. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  5. FBR type reactor

    International Nuclear Information System (INIS)

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  6. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  7. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  8. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  9. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  10. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  11. Comparison of Pickering NGS performance with world power reactors, 1977

    International Nuclear Information System (INIS)

    Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

  12. Heat removal by natural convection in a RPR reactor

    International Nuclear Information System (INIS)

    In this paper natural convection in RPR reactor is analysed. The effect of natural convection valves size on cladding temperature is studied. The reactor channel heat transfer problem is solved using finite elements in a two-dimensional analysis. Results show that two valves with Φ = 0.16 m are suited to keep coolant and cladding temperatures below 730C. (author)

  13. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  14. Helminth Egg Removal Capacity of UASB Reactors under Subtropical Conditions

    NARCIS (Netherlands)

    Yaya-Beas, R.E.; Ayala-Limaylla, C.; Kujawa-Roeleveld, K.; Van Lier, J.B.; Zeeman, G.

    2015-01-01

    This research was conducted to study the anaerobic sludge filtration capacity regarding helminth egg removal in upflow anaerobic sludge blanket (UASB) reactors. Two 25 L lab-scale UASB reactors were operated at an ambient temperature which varied between 17.1 and 28.6 °C. Ascaris suum egg was select

  15. Multi-purpose reactor

    International Nuclear Information System (INIS)

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MWth, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co60) production capacity is 50000 Ci/yr, 200Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  16. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author)

  17. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  18. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  19. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  20. Pebble bed modular reactor (PBMR)

    International Nuclear Information System (INIS)

    In 1993, the pebble bed modular reactor (PBMR) was identified by ESKOM, the electric utility of South Africa, as a leading option for the installation of new generating capacity to their electric grid. This innovative nuclear power plant incorporates a closed cycle primary coolant system utilizing helium to transport heat energy directly from the modular pebble bed reactor to a recuperative power conversion unit with a single-shaft turbine/compressor/generator. This replacement of the steam cycle that is common in present nuclear power plants (NPP) with a direct gas cycle provides the benefits of simplification and a substantial increase in overall system efficiency with the attendant lowering of capital and operational costs. Although the historical development of this plant is interrelated to other types of high temperature gas cooled reactors (HTGRs), the principle focus herein is on the pebble bed (spherical) fuel element type reactor. The long-term development of this reactor type began in Germany by the KFA Nuclear Research Center (now FZJ). Two pebble bed plants were constructed in Germany, the 46 MW(th)/15 MW(e) Arbeitsgemeinshaft Versuchsreaktor (AVR) and the 750 MW(th)/296 MW(e) thorium high temperature reactor (THTR-300). Basically, these steam/electric plants validated the temperature and fission product retention capabilities of the ceramic (TRISO) coated fuel particle and the safety characteristics of the HTGR. Most notable of the operational achievements was with the AVR in sustaining longterm operation at an average core outlet temperature of 950 deg. C, and in demonstration of safety such as extended loss of forced cooling on the core. More details on the AVR and THTR-300 plants are provided The next evolution of the pebble bed plant began in the early 1980s with development of the modular reactor. This small reactor added the unique characteristic of being able to cool the core entirely by passive heat transfer mechanisms following postulated

  1. New research reactor for Australia

    International Nuclear Information System (INIS)

    HIFAR, Australia's major research reactor was commissioned in 1958 to test materials for an envisaged indigenous nuclear power industry. HIFAR is a Dido type reactor which is operated at 10 MW. With the decision in the early 1970's not to proceed to nuclear power, HIFAR was adapted to other uses and has served Australia well as a base for national nuclear competence; as a national facility for neutron scattering/beam research; as a source of radioisotopes for medical diagnosis and treatment; and as a source of export revenue from the neutron transmutation doping of silicon for the semiconductor industry. However, all of HIFAR's capabilities are becoming less than optimum by world and regional standards. Neutron beam facilities have been overtaken on the world scene by research reactors with increased neutron fluxes, cold sources, and improved beams and neutron guides. Radioisotope production capabilities, while adequate to meet Australia's needs, cannot be easily expanded to tap the growing world market in radiopharmaceuticals. Similarly, neutron transmutation doped silicon production, and export income from it, is limited at a time when the world market for this material is expanding. ANSTO has therefore embarked on a program to replace HIFAR with a new multi-purpose national facility for nuclear research and technology in the form of a reactor: a) for neutron beam research, - with a peak thermal flux of the order of three times higher than that from HIFAR, - with a cold neutron source, guides and beam hall, b) that has radioisotope production facilities that are as good as, or better than, those in HIFAR, c) that maximizes the potential for commercial irradiations to offset facility operating costs, d) that maximizes flexibility to accommodate variations in user requirements during the life of the facility. ANSTO's case for the new research reactor received significant support earlier this month with the tabling in Parliament of a report by the Australian Science

  2. Viability of resistance thermometers under reactor conditions

    International Nuclear Information System (INIS)

    The viability of six commercial resistance thermometers is examined under reactor conditions; the sensitive element was P1-2 platinum wire of diameter 50 /mu/m. The resistances R/sub 0/ of the thermometers at 0/degree/C were 46.00/plus or minus/0.02 /OMEGA/while the ratios were R/sub 100//R/sub 0/.1.391/plus or minus/0.0007. The design and manufacturing technology have been described. The thermometers were irradiated in a VVR-M reactor. The thermometers were calibrated before and after irradiation outside the reactor by the reference-point method at temperatures of l00, 231.8, and 327.2/degree/C. 7 refs

  3. FUSED REACTOR FUELS

    Science.gov (United States)

    Mayer, S.W.

    1962-11-13

    This invention relates to a nuciear reactor fuel composition comprising (1) from about 0.01 to about 50 wt.% based on the total weight of said composition of at least one element selected from the class consisting of uranium, thorium, and plutonium, wherein said eiement is present in the form of at least one component selected from the class consisting of oxides, halides, and salts of oxygenated anions, with components comprising (2) at least one member selected from the class consisting of (a) sulfur, wherein the sulfur is in the form of at least one entity selected irom the class consisting of oxides of sulfur, metal sulfates, metal sulfites, metal halosulfonates, and acids of sulfur, (b) halogen, wherein said halogen is in the form of at least one compound selected from the class of metal halides, metal halosulfonates, and metal halophosphates, (c) phosphorus, wherein said phosphorus is in the form of at least one constituent selected from the class consisting of oxides of phosphorus, metal phosphates, metal phosphites, and metal halophosphates, (d) at least one oxide of a member selected from the class consisting of a metal and a metalloid wherein said oxide is free from an oxide of said element in (1); wherein the amount of at least one member selected from the class consisting of halogen and sulfur is at least about one at.% based on the amount of the sum of said sulfur, halogen, and phosphorus atom in said composition; and wherein the amount of said 2(a), 2(b) and 2(c) components in said composition which are free from said elements of uranium, thorium, arid plutonium, is at least about 60 wt.% based on the combined weight of the components of said composition which are free from said elements of uranium, thorium, and plutonium. (AEC)

  4. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  5. Design and fabrication of HTTR reactor pressure vessel

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is under construction at the Oarai Research Establishment, JAERI and planned to be critical at the end of 1997. The HTTR is a High Temperature Gas-cooled Reactors (HTGRs) with thermal output of 30MW, inlet coolant temperature of 395degC, and outlet coolant temperature of 850degC at rated operation and 950degC at high temperature test operation. 2.25Cr-1Mo steel is chosen for the reactor pressure vessel of the HTTR because its temperature reaches about 400degC at normal operation. 2.25Cr-1Mo steel has higher creep rupture strength than Mn-Mo steel used for the reactor pressure vessels of Light Water Reactors (LWRs). For the components of the HTTR reactor pressure vessel subjected to low temperatures where creep deformation is negligible, a design guideline based on Japanese structural design standard for LWRs 'Technical standards for LWR power plant components-Ministry of International Trade and Industry Standard No.501' is utilized. On the other hand, design of the components for high temperature application, where creep behavior dominates, is conducted under newly determined high temperature structural design guideline and design material data. The fabrication of the HTTR reactor pressure vessel took about 23 months. It was installed in a reactor containment vessel in August, 1994. After core components had been installed in the reactor pressure vessel, pressure test of the primary and secondary cooling system including the reactor pressure vessel was performed and successfully ended in March, 1996. This paper reports issues of the HTTR reactor pressure vessel such as structure, material, stress analysis, fabrication, examination and testing. (author)

  6. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

    In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

  7. Model Based Cyber Security Analysis for Research Reactor Protection System

    International Nuclear Information System (INIS)

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN

  8. Status of Jordan Research and Training Reactor Project

    International Nuclear Information System (INIS)

    The Jordan Atomic Energy Commission launched a construction project of Jordan's first nuclear reactor in 2010, following a contract with the consortium of Korea Atomic Energy Research Institute and Daewoo E and C. The Jordan Research and Training Reactor (JRTR), a 5 MW multipurpose research reactor, will serve as the focal point for a national nuclear technology center. This article presents the status of the JRTR Project and future plan as well. With the Construction Permit issued by the Jordan Nuclear Regulatory Commission, the construction work of reactor building and service building is underway, while over 95% of design work has been finished

  9. An overview of third generation reactors - Presentation

    International Nuclear Information System (INIS)

    Now, in 2013 about 69 nuclear reactors (67 GWe) are being built in the world, mostly in China and Russia. Although a few second generation reactors are being built (particularly in China), third generation reactors seem to be systematically chosen for any new construction. The French commercial offer is based on 3 models: the EPR, the ATMEA-1 and the KERENA, the first 2 being of PWR-type while the latter is a BWR. A lot of third generation reactor designs are available on the international market. Concerning the PWR technology we have: the EPR (AREVA), the AP1000 (Westinghouse), the AES 2006 and the VVER TOI (Rosatom), the APR1400 (KEPCO), the APWR (MHI), the ATMEA-1 (AREVA + MHI), the ACP 1000 (CNNC - China), ACPR1000 (CGN - China) and CAP 1400 (SNPTC - China). Concerning the BWR technology, the commercial offer is far less important we have: the ABWR (GE-Hitachi + Toshiba) and the ESBWR (GE-Hitachi). Some third generation reactors are operating now: the first ABWR and AES reactors but most of them (EPR, AP1000, APR1400 and AES 2006) are at an advanced stage of construction. (A.C.)

  10. Gas-cooled reactors and their applications

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review and discuss the current status and recent progress made in the technology and design of gas-cooled reactors and their application for electricity generation, process steam and process heat production. The meeting was attended by more than 200 participants from 25 countries and International Organizations presenting 34 papers. The technical part of the meeting was subdivided into 7 sessions: A. Overview of the Status of Gas-Cooled Reactors and Their Prospects (2 papers); B. Experience with Gas-Cooled Reactors (5 papers); C. Description of Current GCR Plant Designs (10 papers); D. Safety Aspects (4 papers); E. Gas-Cooled Reactor Applications (3 papers); F. Gas-Cooled Reactor Technology (6 papers); G. User's Perspectives on Gas-Cooled Reactors (4 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. A separate abstract was prepared for each of the 34 presentations of this meeting. Refs, figs and tabs

  11. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  12. Safety of research reactors

    International Nuclear Information System (INIS)

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  13. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  14. Treatment of domestic wastewater in an up-flow anaerobic sludge blanket reactor followed by moving bed biofilm reactor

    NARCIS (Netherlands)

    Tawfik, A.; El-Gohary, F.; Temmink, B.G.

    2010-01-01

    The performance of a laboratory-scale sewage treatment system composed of an up-flow anaerobic sludge blanket (UASB) reactor and a moving bed biofilm reactor (MBBR) at a temperature of (22-35 A degrees C) was evaluated. The entire treatment system was operated at different hydraulic retention times

  15. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  16. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  17. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

  18. The economics of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Specific technical features of fast reactors make them more expensive than water-cooled reactors in terms of initial investment, an over cost of 30% is acknowledged in this study. Their consumption of natural uranium is negligible being fed on depleted uranium (except for the very first cycle when an important quantity of plutonium is necessary). In the context of the scarcity of natural uranium, fast reactors could provide a competitive KWh compared with PWR. The study shows that sodium-cooled fast reactor could be economically competitive somewhere in the second part of the 21. century. The development of fast reactors could be accelerated by other arguments than economic competitiveness, for instance some governments might value more the energy independence given by a fleet of fast reactors or by considerations linked to non-proliferation or to the burning of actinides. In addition the article details the worldwide resource in natural uranium. (A.C.)

  19. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  20. FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  1. Reactor water sampling device

    International Nuclear Information System (INIS)

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  2. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  3. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  4. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  5. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  6. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  7. Reactor de plasma

    OpenAIRE

    Erra Serrabasa, Pilar; Molina Mansilla, Ricardo; Beltrán Serra, Eric

    2008-01-01

    Reactor de plasma. Se trata de un reactor de plasma que puede trabajar en un amplio rango de presión, desde el vacío y presiones reducidas hasta la presión atmosférica y presiones superiores. Adicionalmente el reactor de plasma tiene la capacidad de regular otros parámetros importantes y permite su uso para el tratamiento de muestras de tipología muy diversa, como por ejemplo las de tamaño relativamente grande o de superficie rugosa.

  8. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  9. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  10. Licensed operating reactors

    International Nuclear Information System (INIS)

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  11. First Algerian research reactor

    International Nuclear Information System (INIS)

    In 1985, both the Algerian Commissariat of New Energies and the Argentine National Atomic Energy Commission plus the firm INVAP S.E., started a series of mutual visits aimed at defining the mechanisms for cooperation in the nuclear field. Within this framework, a commercial contract was undersigned covering the supply of a low-power reactor (RUN), designed for basic and applied research in the fields of reactor physics and nuclear engineering. The reactor may also be used for performing experiences with neutron beams, for the irradiation of several materials and for the training of technicians, scientists and operators

  12. Course on reactor physics

    International Nuclear Information System (INIS)

    In Germany only few students graduate in nuclear technology, therefore the NPP operating companies are forced to develop their own education and training concepts. AREVA NP has started together with the Technical University of Dresden a one-week course ''reactor physics'' that includes the know-how of the nuclear power plant construction company. The Technical University of Dresden has the training reactor AKR-2 that is retrofitted by modern digital instrumentation and control technology that allows the practical training of reactor control.

  13. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  14. Nuclear reactor theory

    International Nuclear Information System (INIS)

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  15. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  16. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  17. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  18. Joyo experimental reactor tour

    International Nuclear Information System (INIS)

    JAEA cooperation in remote monitoring focuses on the Joyo Experimental Reactor at the O'arai Research and Development Center. Joyo performs irradiation of test fuels to support development of the fast reactor cycle in Japan, both in international cooperation and in support of the Monju fast reactor, which is now undergoing reconstruction. The tour included an introduction at the model, a visit to the control room, entry into the containment vessel, and viewing of remote monitoring equipment in the Fresh Fuel Storage and at one of the Spent Fuel Ponds. (author)

  19. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  20. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  1. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

  2. Reactor monitoring system

    International Nuclear Information System (INIS)

    The present invention concerns a device for monitoring the inside of an FBR type reactor which can not be monitored by a usual optical camera. An ultrasonic camera having an excellent propagating property in a liquid metal sodium is scanned, and reflected waves of the ultrasonic waves are received as signals. The signals are processed by using a virtual realistic feeling (VR) technique such as a head mounting type image display (HMD) and a three dimensional pointing device. With such procedures, the inside of the FBR type reactor can be observed with such a realistic feeling that the inside of the FBR type reactor were seen directly. (I.S.)

  3. Research reactor support

    International Nuclear Information System (INIS)

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  4. Dating of the Francevillian sedimentary series and mineralogic and isotopic (Sm, Nd, Rb, Sr, K, Ar, U, O and C) characterization of the gangue of the reactors 10 and 13. Preliminary report

    International Nuclear Information System (INIS)

    This paper summarizes the various ages reported for the diagenetic events in the Francevillian sedimentary series (Precambrian era) and the fission reactors of Oklo. Obviously, differences exist between the ages obtained on the silicate minerals and the ages obtained on the Uranium ores and on the reactors. Clay minerals which crystallized during the fission reactions yield younger ages than the reactors themselves. Similarly, the diagenetic clays (1870 Ma) show younger ages than the Uranium ores (2000 Ma). This is in contrast to mineralogical and field evidence indicating that Uranium mineralization occurred during diagenesis of the Francevillian sediments. These antithetical results give rise to several questions. Does the age obtained on the diagenetic clays date a late thermal event or does the age of the Uranium mineralization reflect a multistage U-Pb history. This work tries to bring answers with the help of new isotopic analysis and studies mineralogy of the gangue of reactors and isotopic compositions in Uranium ores. 8 refs., 4 figs

  5. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    International Nuclear Information System (INIS)

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I ampersand C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models

  6. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  7. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... nuclear power or research reactor in the United States: Austria Belgium Bulgaria Canada Czech Republic... research reactor in those countries, or (3) The component is in semifabricated form and will be undergoing... form, for research reactors capable of continuous operation above 5 MWe thermal. (c) This...

  8. 8500C VHTR plant technical description

    International Nuclear Information System (INIS)

    This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 8500C (15620F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor design is similar to a previously reported VHTR design having a 9500C (17420F) core outlet temperature

  9. Study of power reactor dynamics by stochastic reactor oscillator method

    International Nuclear Information System (INIS)

    Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber

  10. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  11. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention concerns a method of controlling reactor power to shift it into a partial power operation upon occurrence of recycling pump tripping or loss of generator load. Operation state of a reactor is classified into a plurality of operation states based on values of the reactor core flow rate and the reactor power. Different insertion patterns for selected control rods are determined on every classified operation states. Then, an insertion pattern corresponding to the operation state upon occurrence of recycling pump tripping or loss of power generator load is carried out to shift into partial power operation. The operation is shifted to a load operation solely in the station while avoiding risks such as TPM scram. Then neutron fluxes are suppressed upon transient to increase margin of fuel integrity. Selected control rod pattern of the optimum reactivity is set to each of operation regions, thereby enabling to conduct flexible countermeasure so as to attain optimum operationability. (N.H.)

  12. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  13. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  14. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  15. Reactor parameter simulation system

    International Nuclear Information System (INIS)

    A reactor parameter simulation system (RPSS) has been built with the capability of analyzing any reactor signals, decomposing those signals into their deterministic and stochastic components, then reconstructing new, simulated signals that possess the same statistical and correlation structure as the original plant variables. Important uses of the RPSS are for integration with reactor simulation software to provide tools for plant control strategy development, and for safety-study investigations of scenarios that can arise involving signal faults generated from degraded sensors. A third use of the RPSS is for frequency-domain filtering of reactor process variables contaminated with serially correlated noise, which is important for our ongoing development of expert systems for sensor-operability surveillance. 5 refs., 4 figs., 3 tabs

  16. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  17. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  18. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  19. Ageing of research reactors

    International Nuclear Information System (INIS)

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  20. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  1. Dossier: research reactors

    International Nuclear Information System (INIS)

    Research reactors are used at the CEA (the French atomic energy commission) since many years. Their number has been reduced but they remain unique tools that CEA valorize continuously. The results of the programs involving such reactors are of prime importance for the operation of Electricite de France (EdF) park of existing power plants but also for the design of future nuclear power plants and future research reactors. This dossier presents three examples of research reactors in use at the CEA: Osiris and Orphee (CEA-Saclay), devoted to nuclear energy and fundamental research, respectively, and the critical mockups Eole, Minerve and Masurca (CEA-Cadarache) devoted to nuclear data libraries and neutronic calculation. (J.S.)

  2. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  3. Future Reactor Experiments

    CERN Document Server

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measurement techniques have been explored. A proposed experiment JUNO, with a 20 kton liquid scintillator detector of $3%/$$\\sqrt{E(MeV)}$ energy resolution, $\\sim$ 53 km far from reactors of $\\sim$ 36 GW total thermal power, can reach to a sensitivity of $\\Delta\\chi^{2}>16$ considering the spread of reactor cores and uncertainties of the detector response. Three of mixing parameters are expected to be measured to better than 1% precision. There are multiple detector options for JUNO under investigation. The technical challenges...

  4. Reactor hot spot analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  5. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  6. Research Reactor Benchmarks

    International Nuclear Information System (INIS)

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given

  7. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author)

  8. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  9. Pulsed fusion reactors

    International Nuclear Information System (INIS)

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

  10. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  11. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  13. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  14. The replacement research reactor

    International Nuclear Information System (INIS)

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  15. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  16. Small reactor return

    International Nuclear Information System (INIS)

    Current state of the development of present-day small reactors in different countries is performed. Various designs of low and middle power reactors, among which are CAREM (25 MW, PWR), KLT-40 (40 MW, PWR), MRX (30 MW, PWR), IRIS (50 MW, PWR), SMART (1000 MW, PWR), Modular SBWR (50 MW, BWR), PBMR (120 MW, HTGR), GT-HMR (285 MW, HTGR), are discussed

  17. Reactor lattice transport calculations

    International Nuclear Information System (INIS)

    The present lecture is a continuation of the lecture on Introduction to the Neutron Transport Phenomena. It comprises three aspects of lattice calculations. First the idea of a reactor lattice is introduced. Then the main definitions used in reactor lattice analysis are given, and finally two basic methods applied for solution of the transport equations are defined. Several remarks on secondary results from lattice transport calculations are added. (author)

  18. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  19. Future reactor experiments

    International Nuclear Information System (INIS)

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper

  20. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them