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Sample records for c reactor

  1. New inlet nozzle assembly: C Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Calkin, J.F.

    1960-10-19

    The use of self-supported fuel elements in ribless Zircaloy-2 tubes at C-Reactor requires some inlet nozzle modification to allow charging of the larger overall diameter fuel pieces. A new nozzle assembly has been developed (by Equipment Development Operation -- IPD) which will allow use of the new fuel pieces and at the same time increase the reliability of the header-to-tube piping and reduce pumping power losses. Flow test data were requested for the new assembly and the results of these tests are presented herein. This report also presents a comparison of the header to tube energy losses for the various reactor inlet nozzle assemblies which are currently used on the Hanford production reactors.

  2. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

  3. 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.

  4. 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

  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. 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)

  7. 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

  8. Irradiation Scheme Design of 14C Production on 49-2 Reactor

    Institute of Scientific and Technical Information of China (English)

    SUN; Zheng; LIU; Xing-min; XU; Zhi-long; ZHANG; Ya-dong

    2012-01-01

    <正>14C is a radioisotope of carbon, it is widely used in pharmacy, medical treatment, agriculture, reconnoiter and archaeology. 49-2 research reactor is a swimming pool style reactor which has operated for more than 40 years. The application of 49-2 reactor includes the radio nuclides production. Therefore, the technical scheme on 14C irradiation in 49-2 reactor should be prepared elaborately.

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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.)

  16. 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.)

  17. 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.)

  18. 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

  19. Production test PTA-002, increased graphite temperature limit -- B, C and D Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Russell, A.

    1965-12-17

    The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10% uniform oxidation of graphite results in a loss in strength of approximately 50%. Production Test IP-725 was conducted at F Reactor for a period of six months at graphite temperatures approximately 50 and 100 C higher than the present graphite temperature limit of 650 C. The results from the F Reactor test suggest that an increase in the graphite temperature limit from 650 C to 700 C is technically feasible from the standpoint of oxidation of the graphite moderator with CO{sub 2}. Any significant additional increase was shown to lead to excessively high oxidation rates and is therefore not considered feasible. The objective of this test, therefore, is to extend the higher temperature investigations to B, C, and D Reactors. For the duration of this test, the graphite temperature limit will be increased from 650 C and 700 C, corresponding to an increase in the graphite stringer temperature limit from 735 C to 790 C. The test is expected to last for approximately six months but may be terminated early on any or all the reactors.

  20. 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)

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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

  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. 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

  12. 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

  13. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  14. 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.

  15. 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.

  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. 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.

  18. Modeling Forced Flow Chemical Vapor Infiltration Fabrication of SiC-SiC Composites for Advanced Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Christian P. Deck

    2013-01-01

    Full Text Available Silicon carbide fiber/silicon carbide matrix (SiC-SiC composites exhibit remarkable material properties, including high temperature strength and stability under irradiation. These qualities have made SiC-SiC composites extremely desirable for use in advanced nuclear reactor concepts, where higher operating temperatures and longer lives require performance improvements over conventional metal alloys. However, fabrication efficiency advances need to be achieved. SiC composites are typically produced using chemical vapor infiltration (CVI, where gas phase precursors flow into the fiber preform and react to form a solid SiC matrix. Forced flow CVI utilizes a pressure gradient to more effectively transport reactants into the composite, reducing fabrication time. The fabrication parameters must be well understood to ensure that the resulting composite has a high density and good performance. To help optimize this process, a computer model was developed. This model simulates the transport of the SiC precursors, the deposition of SiC matrix on the fiber surfaces, and the effects of byproducts on the process. Critical process parameters, such as the temperature and reactant concentration, were simulated to identify infiltration conditions which maximize composite density while minimizing the fabrication time.

  19. 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)

  20. 14C release from a Soviet-designed pressurized water reactor nuclear power plant.

    Science.gov (United States)

    Uchrin, G; Csaba, E; Hertelendi, E; Ormai, P; Barnabas, I

    1992-12-01

    The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.

  1. 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)

  2. 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

    with household solid waste. Kinetic analysis of the biohydrogen enrichment cultures show that substrate (glucose) likely inhibited hydrogen production when its concentration was higher than 1 g/L. Different start up strategies were applied for biohydrogen production in biofilm reactors operated at 70°C, and fed......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...

  3. 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.

  4. 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

  5. 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

  6. 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 (

  7. 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.

  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. 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

    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......) method, were significantly lower on glucose, acetate and butyrate at the increased operational temperature, while the numbers of hydrogenotrophic methanogens remained unchanged. No viable propionate-degrading bacteria were enriched at 65 degreesC. Use of ribosomal oligonucleotide probes showed...

  10. 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

  11. 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.

  12. 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

  13. Field study of moving bed biofilm reactor technology for post-treatment of wastewater lagoon effluent at 1 degree C.

    Science.gov (United States)

    Almomani, Fares A; Delatolla, Robert; Ormeci, Banu

    2014-08-01

    The goal of this study was to investigate the potential use ofmoving bed biofilm reactor (MBBR) systems as ammonia removal post-treatment units for wastewater (WW) treatment lagoons that demonstrate large temperature changes throughout their operational year (1 - 20 degrees C). The study was carried out over a six-month period using laboratory-scale MBBR reactors fed with incoming effluent from a full-scale lagoon. The study shows that significant average ammonia removal rates of 0.26 and 0.11 kgN/m . d were achieved at 20 degrees C and 1C. The increase in the ammonia removal rates with increasing temperature from 1 degrees C to 20 degrees C showed a strong correlation to an applied temperature correction coefficient model. No significant accumulation of effluent nitrite was observed at 1 degrees C or after being fed with synthetic wastewater (SWW); indicating that cold temperatures and transitions from real WW to SWW did not stress the nitrifiers. Furthermore, the study demonstrates that changes in temperature or changes from real WW to SWW do not affect the mass of biofilm attached per MBBR carrier. Hence, based on the results of this study, it is concluded that MBBR is a promising technology for post-treatment ammonia removal of WW lagoon effluent.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}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{degrees}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 x 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.

  15. 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

  16. Development of C/C composite for the core component of the high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H

    2005-01-15

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite.

  17. 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.

  18. Enrichment of C-14 on Surface Deposits of Oldbury Reactor Graphite Investigated with the Use of Magnetic Sector Secondary Ion Mass Spectrometry

    OpenAIRE

    Payne, Liam; Heard, Peter; Scott, Tom

    2015-01-01

    Large quantities of irradiated graphite containing C-14 will arise from the decommissioning of the UK’s Magnox power stations. Magnetic sector secondary ion mass spectrometry (MS-SIMS) has been used to investigate the distribution of this radioisotope within a number of trepanned graphite samples from a variety of radial and axial positions within reactor one at Oldbury, a Magnox nuclear reactor. The methodology used for these determinations ensured that possible mass interferences between C-...

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. Comparison of polyurethane foam and biodegradable polymer as carriers in moving bed biofilm reactor for treating wastewater with a low C/N ratio.

    Science.gov (United States)

    Chu, Libing; Wang, Jianlong

    2011-03-01

    This paper presents a comparison between two different materials used as carriers: inert polyurethane (PU) foam and biodegradable polymer polycaprolactone (PCL) particles for the removal of organics and nitrogen from wastewater with a low C/N ratio using moving bed biofilm reactors. The results, during a monitoring period of four months, showed that TOC and ammonium removal efficiency was higher in reactor 2 filled with PU carriers than in reactor 1 filled with PCL carriers (90% and 65% in the former, compared with 72% and 56% in the latter at an hydraulic retention time of 14 h). Reactor 1 showed good behavior in terms of total nitrogen removal as the biodegradable polymer was an effective substrate providing reducing power for denitrification. From three-dimensional excitation-emission matrix analysis, it was shown that the effluent from reactor 1 contained mainly protein-like and soluble microbial product-like substances.

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

    DEFF Research Database (Denmark)

    Zhang, Long; Liang, Shanshan; Hellgren, Lars;

    2008-01-01

    of sphingomyelin has been proven to be a feasible method to produce ceramide. In the membrane reactor constructed, the aqueous phase and the organic phase were separated by a membrane containing the immobilized enzyme, while the organic phasewas continuously circulated. Among the 10 selected membranes, the enzyme...... immobilized in membrane RC 70PP had low immobilization efficiency, but retained the highest catalytic activity. Three immobilization methods, i.e. filtration (adsorption/entrapment), covalent binding, and cross-linking, were compared. The enzyme immobilized by filtration had the highest activity even under...

  5. Nitrifying moving bed biofilm reactor (MBBR) biofilm and biomass response to long term exposure to 1 °C.

    Science.gov (United States)

    Hoang, V; Delatolla, R; Abujamel, T; Mottawea, W; Gadbois, A; Laflamme, E; Stintzi, A

    2014-02-01

    This study aims to investigate moving bed biofilm reactor (MBBR) nitrification rates, nitrifying biofilm morphology, biomass viability as well as bacterial community shifts during long-term exposure to 1 °C. Long-term exposure to 1 °C is the key operational condition for potential ammonia removal upgrade units to numerous northern region treatment systems. The average laboratory MBBR ammonia removal rate after long-term exposure to 1 °C was measured to be 18 ± 5.1% as compared to the average removal rate at 20 °C. Biofilm morphology and specifically the thickness along with biomass viability at various depths in the biofilm were investigated using variable pressure electron scanning microscope (VPSEM) imaging and confocal laser scanning microscope (CLSM) imaging in combination with viability live/dead staining. The biofilm thickness along with the number of viable cells showed significant increases after long-term exposure to 1 °C. Hence, this study observed nitrifying bacteria with higher activities at warm temperatures and a slightly greater quantity of nitrifying bacteria with lower activities at cold temperatures in nitrifying MBBR biofilms. Using DNA sequencing analysis, Nitrosomonas and Nitrosospira (ammonia oxidizers) as well as Nitrospira (nitrite oxidizer) were identified and no population shift was observed between 20 °C and after long-term exposure to 1 °C.

  6. 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.

  7. 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

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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...

  15. 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.

  16. 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)

  17. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jiang, Weilin [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    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.

  19. 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.)

  20. 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 ...

  1. 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

  2. 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.

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

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    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.

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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.

  10. Characteristics of biofilm attaching to carriers in moving bed biofilm reactor used to treat vitamin C wastewater.

    Science.gov (United States)

    Hu, Xiao-bing; Xu, Ke; Wang, Zhao; Ding, Li-li; Ren, Hong-qiang

    2013-01-01

    In order to investigate characteristics of biofilm attaching firmly to carriers in the moving bed biofilm reactor (MBBR) used for vitamin C wastewater treatment, experiments were undertaken with instrumental analysis methods. Scanning electron microscopy (SEM) micrographs of MBBR biofilms revealed that there were rod-shaped microbes and cocci in the biofilm, and microbes were embedded within medium substances and the biofilm matrix adhered firmly to carriers, leading to the formation of a smooth compacted surface at the base of the biofilm. Transmission electron microscopy (TEM) analysis revealed that extracellular polymeric substances (EPS) layer surrounded cell, sequestered inorganics to form a mixed structure, which ensured firm attachment of the biofilm to the carrier. X-ray diffraction (XRD) experiments and thermogravimetry analysis revealed that (i) the biofilm contained many inorganic substances, about 70.5%, and the inorganic substances contained multiple classes of inorganic with a high boiling point; (ii) inorganic elements such as calcium and phosphorous were selectively absorbed and accumulated in the biofilm as insoluble compounds with amorphous phases, rendering the biofilm highly resistant to detachment. Fourier-transform infrared (FTIR) spectroscopy showed carbohydrates were the main EPS.

  11. 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

  12. 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.

  13. 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...

  14. 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.

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

    Science.gov (United States)

    Pauzi, A. M.

    2013-06-01

    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.

  16. 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...

  17. 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

  18. 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 ...

  19. 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

  20. 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

  1. Determination of {beta}{sub eff} using MCNP-4C2 and application to the CROCUS and PROTEUS reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollaire, J. [European Organization for Nuclear Research CERN, CH-1211 Geneve 23 (Switzerland); Plaschy, M.; Jatuff, F. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, CH-1015 Lausanne (Switzerland)

    2006-07-01

    A new Monte Carlo method for the determination of {beta}{sub eff} 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 {beta}{sub eff} per delayed neutron group, per fissioning nuclide and per reactor region is illustrated. Finally, the MCNP predictions of {beta}{sub eff} are compared to the results of deterministic calculations. (authors)

  2. 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.

  3. 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...

  4. 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

  5. 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

  6. 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

  7. 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)

  8. Bistability in isothermal photochemical systems: The A ⇆ h nu B --> h nu C reaction in a continuous flow stirred tank reactor

    Science.gov (United States)

    Laplante, J. P.; Lavabre, D.; Micheau, J. C.

    1988-08-01

    In this paper we present a kinetic analysis of the consecutive photoreaction scheme A⇄hνB→hνC assuming the reaction is carried out in a continuous flow stirred tank reactor (CSTR). The reactor is kept at constant temperature and fed with reactant A at a constant flow rate. A numerical analysis of the model's stationary states reveals a range of constraints for which the system possesses multiple steady states. The observed bistability depends strongly on the rate constant of the B→A reaction k2 . It is typically observed when k2 is much larger than the other rate constants. Our numerical calculations also reveal a marked dependency on parameters such as the molar absorptivities and the irradiation intensity I0 . Interestingly, multiple steady states are only observed for intermediate values of I0 . Analytical approximations are obtained for the stationary states in the limit where the end-product C does not absorb light. These approximations are used to clarify the mechanism responsible for the light-induced instability.

  9. A VUV Photoionization Study of the Combustion-Relevant Reaction of the Phenyl Radical (C6H5) with Propylene (C3H6) in a High Temperature Chemical Reactor

    Energy Technology Data Exchange (ETDEWEB)

    University of Hawaii at Manoa; Sandia National Laboratories; Zhang, Fangtong; Kaiser, Ralf I.; Golan, Amir; Ahmed, Musahid; Hansen, Nils

    2012-02-22

    We studied the reaction of phenyl radicals (C6H5) with propylene (C3H6) exploiting a high temperature chemical reactor under combustion-like conditions (300 Torr, 1,200-1,500 K). The reaction products were probed in a supersonic beam by utilizing tunable vacuum ultraviolet (VUV) radiation from the Advanced Light Source and recording the photoionization efficiency (PIE) curves at mass-to-charge ratios of m/z = 118 (C9H10+) and m/z = 104 (C8H8+). Our results suggest that the methyl and atomic hydrogen losses are the two major reaction pathways with branching ratios of 86 10 percent and 14 10 percent. The isomer distributions were probed by fitting the recorded PIE curves with a linear combination of the PIE curves of the individual C9H10 and C8H8 isomers. Styrene (C6H5C2H3) was found to be the exclusive product contributing to m/z = 104 (C8H8+), whereas 3-phenylpropene, cis-1-phenylpropene, and 2-phenylpropene with branching ratios of 96 4 percent, 3 3 percent, and 1 1 percent could account for signal at m/z = 118 (C9H10+). Although searched for carefully, no evidence of the bicyclic indane molecule could be provided. The reaction mechanisms and branching ratios are explained in terms of electronic structure calculations nicely agreeing with a recent crossed molecular beam study on this system.

  10. 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)

  11. 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.)

  12. 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)

  13. 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

  14. Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

    Energy Technology Data Exchange (ETDEWEB)

    Huda, M.Q. [Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, GPO Box 3787, Dhaka-1000 (Bangladesh)]. E-mail: quamrul@dhaka.net

    2006-08-15

    The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for {sup nat}Zr, {sup nat}Mo, {sup nat}Cr, {sup nat}Fe, {sup nat}Ni, {sup nat}Si, and {sup nat}Mg) at 300 K evaluations were used. Full S({alpha}, {beta}) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is {approx}20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: {delta} {sup 28}, {delta} {sup 25}, {rho} {sup 25}, and C {sup *} were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.

  15. 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.

  16. The use and evolution of the CEA research reactors; Utilisation et evolution des reacteurs de recherche du C.E.A

    Energy Technology Data Exchange (ETDEWEB)

    Rossillon, F.; Chauvez, C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The authors successively examine the different research reactors in use in the French C.E.A. Nuclear Centres. They trace briefly their histories, describing how they have been used up to the present, and how they have been adapted to changes in programme by means of certain modifications. They also describe the reasons which have led to the elaboration of the project for the new reactor Osiris. Zoe, the oldest reactor in the CEA, has been in service in the Centre de Fontenay-aux-Roses since 1948. It is used mainly for measurements of absorption cross-sections in graphite, and for various short irradiations which do not require high fluxes. The reactor EL 2, in service since 1952, was used for the first studies on gas cooling. It has also been widely used for the production of radioisotopes and for a large number of experiments in the fields of physics, metallurgy and physical chemistry. The ageing of certain elements of the reactor has led to the decision to close it down in the near future The reactor EL 3 has been widely used for experiments in physics and in the investigation of fuels. The possibilities of the reactor in fast neutron irradiations will be considerably improved by the adoption of a new type of core (the 'snow crystal' structure). Triton-I, a 2 MW swimming-pool reactor, is used for the most part for fast neutron and gamma irradiations. The modifications being carried out on it at present should result in an increase in the power of the reactor up to 4 or 5 MW. In a neighbouring compartment is housed Triton-II which is of the same general structure, as Triton-I, but whose maximum power is 100 kW. Triton-II is used solely for studies on shielding. Melusine, a 2 MW swimming-pool reactor, has been in use in the Centre d'Etudes Nucleaires de Grenoble since 1959. It has supported a very high programme concerned mainly with solid state physics, fundamental research into refractory fissile materials and special graphites, and the study of

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); 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 [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Borlaug, Brennan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jiang, Weilin [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    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.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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)

  5. 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

  6. Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

    2013-11-15

    Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

  7. 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)

  8. Rapid analysis of 14C and 3H in graphite and concrete for decommissioning of nuclear reactor

    DEFF Research Database (Denmark)

    Hou, Xiaolin

    2005-01-01

    A rapid oxidizing combustion method using a commercial Sample Oxidizer has been investigated to determine separately the C-14 and H-3 activities in graphite and concrete. By this method the sample preparation time can be reduced to 2-3min. The detection limits for H-3 and C-14 are 0.96 and 0.58Bq....../g graphite and 0.11 and 0.06Bq/g concrete, respectively. The cross contamination of C-14 and tritium in the preparation of samples is less than 0.2%. The interference of other radionuclides in the determination of C-14 and tritium in graphite is insignificant. The analytical accuracy, investigated...... by the standard addition method, is better than 95%. In addition, an acid digestion method has also been used to examine the graphite and concrete activities, to allow comparison with the method developed herein. The two methods show good agreement for graphite samples. Graphite samples were collected from...

  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. 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)

  11. 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)

  12. 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.

  13. 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)

  14. Status quo of the storage of C-14 containing waste from German high-temperature and research reactors and the strategy for their future storage; Status quo der Lagerung C-14-haltiger Abfaelle aus deutschen Hochtemperatur- und Forschungsreaktoren und Strategie zu deren kuenftigen Endlagerung. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Doerr, S.; Bollingerfehr, W.; Filbert, W.

    2014-12-15

    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.

  15. Perspectives of SiC-Based Ceramic Composites and Their Applications to Fusion Reactors 6.Recent Research Activities regarding SiC-Based Ceramic Composites for Aerospace Applications

    Science.gov (United States)

    Ogasawara, Toshio

    In this article, the present and future prospects of the research and development regarding continuous SiC fiber reinforced ceramic matrix composites (CMCs) for aerospace applications are reviewed. These activities in Japan are described in term of their major applications, i.e. turbo fan engine components for aircrafts, rocket propulsion components, thermal protection system for future re-entry vehicles, thruster for satellites. It is suggested that high performance, affordable processing cost, and excellent reliability will be important factors in the practical use of CMCs in the future.

  16. 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.)

  17. 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

  18. 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.

  19. 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.

  20. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  1. 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)

  2. 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.)

  3. 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)

  4. 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)

  5. Application of LY-C2-12 D catalyst in C2 hydrogenation isothermal reactor of 330 kt·a-1 ethylene plant%LY-C2-12 D催化剂在330 kt · a-1乙烯装置碳二加氢等温反应器的工业应用

    Institute of Scientific and Technical Information of China (English)

    赵玉龙

    2015-01-01

    中国石油大庆石化公司330 kt·a-1乙烯装置是我国较早成套引进的乙烯装置。该装置碳二加氢一段反应器采用等温反应器,列管之间使用丁烷为冷却介质。第一段反应器采用新型LY-C2-12 D催化剂,运转初期乙烯选择性为80%,运行6个多月,平均选择性达70%,表明该催化剂具有良好的稳定性。%330 kt·a-1 ethylene plant in PetroChina Daqing Petrochemical Co. is one of the earliest im-ported ethylene plants in China. The isothermal reactor is applied as C2 hydrogenation reactor and butane is used for cooling medium among the tubes. The new LY-C2-12D catalyst was applied in the first stage reactor. The selectivity to ethylene reached 80% in initial running period,and the average selectivity was up to 70%,which indicated that the catalyst possessed good stability.

  6. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    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...... microorganism into methane. In thermophilic biogas reactors,, acetate oxidizing cultures occupied the niche of Methanothrix species, aceticlastic methanogens which dominate at low acetate concentrations in mesophilic systems. Normally, thermophilic biogas reactors are operated at temperatures from 52 to 560 C....... Experiments using biogas reactors fed with cow manure showed that the same biogas yield found at 550 C could be obtained at 610 C after a long adaptation period. However, propionate degradation was inhibited by increasing the temperature....

  7. M and c'99 : Mathematics and computation, reactor physics and environmental analysis in nuclear applications, Madrid, September 27-30, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J. M.; Ahnert, C.; Cabellos, O.

    1999-07-01

    The international conference on mathematics and computation, reactor physics and environmental analysis in nuclear applications in the biennial topical meeting of the mathematics and computation division of the American Nuclear Society. (Author)

  8. 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.

  9. 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.

  10. 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

  11. 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...

  12. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  13. 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

    2014-12-01

    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. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  14. 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

  15. 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.)

  16. 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.)

  17. 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.

  18. C

    NARCIS (Netherlands)

    Sinninghe Damsté, J.S.; Schouten, S.; Volkman, J.K.

    2014-01-01

    A limited suite of C-27, C-29 and C-30 rearranged hopenes identified as neohop-13(18)-enes have been reported in immature Recent and ancient marine/lacustrine sediments and their presence has been explained by dehydration and isomerisation of ubiquitous hopanols or hopenes. Here we investigated the

  19. 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...

  20. 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.)

  1. 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

  2. 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

  3. Study on the Adaptability of Etheriifcation Feedstock to Reactor Type

    Institute of Scientific and Technical Information of China (English)

    Mao Junyi; Yuan Qing; Wang Lei; Huang Tao

    2016-01-01

    A reactive C5 oleifns and methanol etheriifcation kinetic model based on E-R mechanism was established and three different types of reactors including the adiabatic ifxed-bed liquid reactor, the external loop reactor and the mixed-phase reactor were constructed by Aspen Plus. The adaptability of reactive C5 oleifns to these reactors was studied and simulated using various gasoline fractions with different oleifns content. After the theoretical model was validated by the experimental data of the etheriifcation of three C5 light cut fractions from different gasoline sources in different reactors, the simulated isoamylene conversion with reactive C5 olefin contents increasing from 10% to 60% was studied in the three different types of reactors for etheriifcation with methanol, respectively. Test results show that there is an obvious adaptability of the feedstock composition to the reactor type to achieve a high conversion.

  4. 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...

  5. 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)

  6. 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.

  7. 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.

  8. 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...

  9. 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)

  10. 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.)

  11. 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.)

  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. 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

  14. 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)

  15. 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.)

  16. 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.)

  17. 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)

  18. 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

  19. Results of charpy V-notch impact testing of structural steel specimens irradiated at {approximately}30{degrees}C to 1 x 10{sup 16} neutrons/cm{sup 2} in a commercial reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Stoller, R.E.

    1997-04-01

    A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at {approximately} 30{degrees}C ({approximately} 85{degrees}F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10{sup 16} neutrons/cm{sup 2} (> 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 {approximately} 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of {approximately} 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.

  20. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    Science.gov (United States)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  1. 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.)

  2. 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)

  3. 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

  4. 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.)

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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.

  10. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    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.

  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. 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)

  13. 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.)

  14. 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.)

  15. 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

  16. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  17. 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)

  18. 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)

  19. 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

  20. 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.)

  1. 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

  2. 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

  3. 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.

  4. On the oxidation of uraninite from natural reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Cui, D.; Eriksen, T.; Eklund, U.B.

    1999-07-01

    Natural nuclear reactors provide unique evidence in helping to understand the processes that might occur over long timescales in radioactive waste disposal sites. In the presented work, the extent and kinetics of oxidation of core material from the Oklo-Bangombe natural reactors are investigated. The X-ray powder diffraction analysis shows that the uraninites core samples from the Bangombe Reactor and Oklo Reactor 2, and Oklo Reactor 13 have the same unit-cell parameters as synthetic UO{sub 2.25}. A significant amount of fourmarierite, Pb(UO{sub 2}){sub 4}O{sub 3}(OH){sub 4}.4H{sub 2}O, was identified in the core samples from two shallow reactors Bangombe and Oklo 2, but not in the deeper reactor Oklo 13. The results of U(IV)/U(IV) measurements indicate that the extent of oxidative weathering of shallow reactors (Bangombe and Oklo 2) is greater than for the deeper reactor Oklo 13. Evaporable organic compounds found in the uraninite inclusion containing bitumen at the edge of Okelobondo Reactor (400 C) and in the black shale immediately above the Bangombe Reactor (260 C) may work as a reducing buffer or/and a hydrophobic water shield to depress the oxidative dissolution of the uraninite cores.

  5. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  6. 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)

  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. 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. 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...

  10. 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

  11. 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.

  12. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    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.

  13. 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

  14. Effects of loading rate and aeration on nitrogen removal and N2O emissions in intermittently aerated sequencing batch reactors treating slaughterhouse wastewater at 11 °C.

    Science.gov (United States)

    Pan, Min; Hu, Zhenhu; Liu, Rui; Zhan, Xinmin

    2015-04-01

    This study aimed to find optimal operation conditions for nitrogen removal from high strength slaughterhouse wastewater at 11 °C using the intermittently aerated sequencing batch reactors (IASBRs) so as to provide an engineering control strategy for the IASBR technology. Two operational parameters were examined: (1) loading rates and (2) aeration rates. Both the two parameters affected variation of DO concentrations in the IASBR operation cycles. It was found that to achieve efficient nitrogen removal via partial nitrification-denitrification (PND), "DO elbow" point must appear at the end of the last aeration period. There was a correlation between the ammonium oxidizing bacteria (AOB)/nitrite oxidizing bacteria (NOB) ratio and the average DO concentrations in the last aeration periods; when the average DO concentrations in the last aeration periods were lower than 4.86 mg/L, AOB became the dominant nitrifier population, which benefited nitrogen removal via PND. Both the nitrogen loading rate and the aeration rate influenced the population sizes of AOB and NOB. To accomplish efficient nitrogen removal via PND, the optimum aeration rate (A, L air/min) applied can be predicted according to the average organic loading rates based on mathematical equations developed in this study. The research shows that the amount of N2O generation in the aeration period was reduced with increasing the aeration rate; however, the highest N2O generation in the non-aeration period was observed at the optimum aeration rates.

  15. 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)

  16. 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.

  17. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  18. 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.

  19. 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.

  20. Microbial degradation of MTBE in reactors

    DEFF Research Database (Denmark)

    Waul, Christopher Kevin; Arvin, Erik; Schmidt, Jens Ejbye

    2007-01-01

    findings were: membrane bioreactors and fluidized bed reactors had the highest volumetric removal rates of all reactors studied, in the order of 1 000 mg/(l d); competition for oxygen, nutrients and occupancy between MTBE degraders and oxidisers of co-contaminants such as, ammonium and the group of benzene......, toluene, ethylbenzene and xylenes, may reduce the removal rates of MTBE, or prevent its removal in reactors. With mathematical modelling, the long startup time required for some MTBE degrading reactors could be predicted. Long startup times of up to 200 days were due to the low maximum growth rate...... of the MTBE degraders, in the order of 0.1 d−1 or less, at 25 °C. However, despite this, high volumetric MTBE removal rates were found to be possible after the startup period when the biomass concentration reached a steady state....

  1. 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

  2. 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

  3. 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.)

  4. 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.)

  5. 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)

  6. 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.

  7. 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

  8. 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

  9. Iris reactor conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V. [Westinghouse Electric Comp., Pittsburgh, PA (United States); Galvin, M.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Lombardi, C.V.; Maldari, F.; Ricotti, M.E. [Politecnico di Milano, Milan (Italy); Cinotti, L. [Ansaldo SpA, Genoa (Italy)

    2001-07-01

    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)

  10. 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.

  11. 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

  12. 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.

  13. 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)

  14. 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)

  15. 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.)

  16. 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.

  17. 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

  18. Design of A solar Thermophilic Anaerobic Reactor for Small Farms

    NARCIS (Netherlands)

    Mashad, El H.; Loon, van W.K.P.; Zeeman, G.; Bot, G.P.A.; Lettinga, G.

    2004-01-01

    A 10 m(3) completely stirred tank reactor has been designed for anaerobic treatment of liquid cow manure under thermophilic conditions (50degreesC), using a solar heating system mounted on the reactor roof. Simulation models for two systems have been developed. The first system consists of loose com

  19. 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.)

  20. 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.)

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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.)

  6. 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.

  7. 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)

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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.

  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. 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.

  15. 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

  16. 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

  17. 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.

  18. 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.)

  19. 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.

  20. 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

  1. Jules Horowitz Reactor: a high performance material testing reactor

    Science.gov (United States)

    Iracane, Daniel; Chaix, Pascal; Alamo, Ana

    2008-04-01

    The physical modelling of materials' behaviour under severe conditions is an indispensable element for developing future fission and fusion systems: screening, design, optimisation, processing, licensing, and lifetime assessment of a new generation of structure materials and fuels, which will withstand high fast neutron flux at high in-service temperatures with the production of elements like helium and hydrogen. JANNUS and other analytical experimental tools are developed for this objective. However, a purely analytical approach is not sufficient: there is a need for flexible experiments integrating higher scales and coupled phenomena and offering high quality measurements; these experiments are performed in material testing reactors (MTR). Moreover, complementary representative experiments are usually performed in prototypes or dedicated facilities such as IFMIF for fusion. Only such a consistent set of tools operating on a wide range of scales, can provide an actual prediction capability. A program such as the development of silicon carbide composites (600-1200 °C) illustrates this multiscale strategy. Facing the long term needs of experimental irradiations and the ageing of present MTRs, it was thought necessary to implement a new generation high performance MTR in Europe for supporting existing and future nuclear reactors. The Jules Horowitz Reactor (JHR) project copes with this context. It is funded by an international consortium and will start operation in 2014. JHR will provide improved performances such as high neutron flux ( 10 n/cm/s above 0.1 MeV) in representative environments (coolant, pressure, temperature) with online monitoring of experimental parameters (including stress and strain control). Experimental devices designing, such as high dpa and small thermal gradients experiments, is now a key objective requiring a broad collaboration to put together present scientific state of art, end-users requirements and advanced instrumentation. To cite this

  2. 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

  3. 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.

  4. 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

  5. NEUTRONIC REACTOR FUEL COMPOSITION

    Science.gov (United States)

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. 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

  12. 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.

  13. 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).

  14. 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.

  15. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  16. 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)

  17. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  18. Gas-solid hydroxyethylation of potato starch in a stirred vibrating fluidized bed reactor

    NARCIS (Netherlands)

    Kuipers, N.J M; Stamhuis, Eize; Beenackers, A.A C M

    1996-01-01

    A novel reactor for modifying cohesive C-powders such as in the gas-solid hydroxyethylation of semidry potato starch is characterized, the so-called stirred vibrating fluidized bed reactor. Good fluidization characteristics are obtained in this reactor for certain combinations of stirring and vibrat

  19. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  20. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  1. OECD Halden reactor project

    International Nuclear Information System (INIS)

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  2. 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.

  3. PUSPATI Triga reactor fuel worth measurement

    International Nuclear Information System (INIS)

    The reactivity worth of fuel elements in the B, C, D, E and F rings in the PUSPATI TRIGA Reactor core with respect to water as well as that of dummy fuel element (graphite filled) in the G ring were measured. The reactivity worth of 8.5 w/o standard TRIGA fuel element with respect to the dummy element in the B to F rings were also determined. The measured results agreed with the typical values given by the reactor supplier, General Atomatic Company, to within eight percents. (author)

  4. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  5. Niobium carbide synthesis by solid-gas reaction using a rotating cylinder reactor

    OpenAIRE

    F. A. O. Fontes; K. K. P. Gomes; de Oliveira, S. A.; Souza, C. P. de; J. F. Sousa

    2004-01-01

    A rotating cylinder reactor was designed for the synthesis of niobium carbide powders at 1173 K. Niobium carbide, NbC, was prepared by carbothermal reduction starting from commercial niobium pentoxide powders. The reactor was heated using a custom-made, two-part, hinged, electric furnace with programmable temperature control. The design and operational details of the reactor are presented. The longitudinal temperature gradient inside the reactor was determined. Total reaction time was monitor...

  6. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  7. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  8. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  9. Innovative energy production in fusion reactors

    International Nuclear Information System (INIS)

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

  10. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Only minor changes were introduced in the French nuclear programme by the new government in 1981. The operating conditions of Rapsodie were very satisfactory up to January 1982. After a leak in the double primary jacket (nitrogen circuit) the reactor was shut down for investigations. Phenix is continuing to operate smoothly. Construction of Super Phenix (Creys Malville power plant) is proceeding normally though with some delay. The studies for the future (after Creys Malville) are following their way both for the Project 1500 (Super Phenix 2) and for the specific plants of the fuel cycle. Research and development are largely directed toward Super Phenix 1 needs and the prospects of Super Phenix 2. International cooperation remains very intensive. The financial resources devoted to the development of fast reactors are globally stable. Including fuel cycle and safety (but excluding the Phenix operation) about 1300 millions of francs will be devoted to fast reactors by the C.E.A. in 1982. (author)

  11. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3He, 6Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  12. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  13. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

  14. High temperature reactor development in the Netherlands

    International Nuclear Information System (INIS)

    This year, some clear design choices have been made in the WHITE Reactor development programme. The activities will be concentrated at the development of a small size pebble bed HTR for combined heat and power production with a closed cycle gas turbine. Objective of the development is threefold: 1. restoring social support; 2. establishing commercial viability after market introduction; and 3. making the market introduction itself feasible, i.e. limited development and first-of-a-kind costs. This design is based on the peu-a-peu design of KFA Juelich and will be optimized. The computer codes necessary for this are being prepared for this work. The dynamic neutronics code PANTHER is being coupled to the thermal hydraulics code THERMIX-DIREKT. For this reactor type, fuel temperatures are maximal in the scenario of depressurization with recriticality. Even for this scenario, fuel temperatures of the 20MWth PAP-GT do not exceed 1300 deg. C, so there should be room for upscaling for economic reasons. On the other hand, it would be convenient to fuel the reactor batchwise instead of continuously, and the use of thorium could be required. These two features may lead to a larger temperature margin. The optimal design must unite these features in the best acceptable way. To gain expertise in calculations on gas cooled graphite moderate reactors, benchmark calculations are being performed in parallel with international partners. Parallel to this, special expertise is being built up on HTR fuel and HTR reactor vessels. (author). 3 refs

  15. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  16. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  17. EBT reactor analysis

    International Nuclear Information System (INIS)

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m2, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density

  18. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  19. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    International Nuclear Information System (INIS)

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  20. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  1. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  2. Methanation assembly using multiple reactors

    Science.gov (United States)

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  3. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  4. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  5. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  6. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report 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 of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  7. 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 of 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

  8. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report 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 of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  9. Reactor safety equipments

    International Nuclear Information System (INIS)

    Purpose: To positively recover radioactive substances discharged in a dry well at the time of failure of a reactor. Constitution: In addition to the emergency gas treating system fitted to a reactor building, a purification system connected through a pipeline to the dry well is arranged in the reactor building. This purification system is connected through pipes fitted to the dry well to forced circulation device, heat exchanger, and purification device. The atmosphere of high pressure steam gases in the dry well is derived to the heat exchanger for cooling, and then radioactive substances which are contained in the gases are removed by filter sets charged with the HEPA filters and the HECA filters. At last, there gases are returned to dry well by circulation pump, repeat this process. (Kamimura, M.)

  10. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report 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 of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  11. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  12. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  13. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  14. The MNSR reactor

    International Nuclear Information System (INIS)

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  15. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  16. Fusion reactor materials

    International Nuclear Information System (INIS)

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

  17. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  18. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  19. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  20. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  1. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  2. Diagnostics for hybrid reactors

    International Nuclear Information System (INIS)

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  3. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  4. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  5. Perspectives on reactor safety

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  6. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  7. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  8. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  9. Reactor gamma spectrometry: status

    International Nuclear Information System (INIS)

    Current work is described for Compton Recoil Gamma-Ray Spectrometry including developments in experimental technique as well as recent reactor spectrometry measurements. The current status of the method is described concerning gamma spectromoetry probe design and response characteristics. Emphasis is given to gamma spectrometry work in US LWR and BR programs. Gamma spectrometry in BR environments are outlined by focussing on start-up plans for the Fast Test Reactor (FTR). Gamma spectrometry results are presented for a LWR pressure vessel mockup in the Poolside Critical Assembly (PCA) at Oak Ridge National Laboratory

  10. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  11. 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

  12. Development of the cascade inertial-confinement-fusion reactor

    International Nuclear Information System (INIS)

    Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li2O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO2 granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation

  13. High-temperature reactor developments in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.).

  14. Risk prevention during reactor shutdown

    International Nuclear Information System (INIS)

    During reactor shutdown potential risks are issued of a number of maintenance operations. In this text we analyse these operations and give the modifications of technical specifications to ameliorate the reactor safety. 4 figs

  15. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  16. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 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. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. 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)

  17. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  18. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  19. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  20. Microstructured reactors for hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Aartun, Ingrid

    2005-07-01

    Small scale hydrogen production by partial oxidation (POX) and oxidative steam reforming (OSR) have been studied over Rh-impregnated microchannel Fecralloy reactors and alumina foams. Trying to establish whether metallic microchannel reactors have special advantages for hydrogen production via catalytic POX or OSR with respect to activity, selectivity and stability was of special interest. The microchannel Fecralloy reactors were oxidised at 1000 deg C to form a {alpha}-Al2O3 layer in the channels in order to enhance the surface area prior to impregnation. Kr-BET measurements showed that the specific surface area after oxidation was approximately 10 times higher than the calculated geometric surface area. Approximately 1 mg Rh was deposited in the channels by impregnation with an aqueous solution of RhCl3. Annular pieces (15 mm o.d.,4 mm i.d., 14 mm length) of extruded {alpha}-Al2O3 foams were impregnated with aqueous solutions of Rh(NO3)3 to obtain 0.01, 0.05 and 0.1 wt.% loadings, as predicted by solution uptake. ICP-AES analyses showed that the actual Rh loadings probably were higher, 0.025, 0.077 and 0.169 wt.% respectively. One of the microchannel Fecralloy reactors and all Al2O3 foams were equipped with a channel to allow for temperature measurement inside the catalytic system. Temperature profiles obtained along the reactor axes show that the metallic microchannel reactor is able to minimize temperature gradients as compared to the alumina foams. At sufficiently high furnace temperature, the gas phase in front of the Rh/Al2O3/Frecralloy microchannel reactor and the 0.025 wt.% Rh/Al2O3 foams ignites. Gas phase ignition leads to lower syngas selectivity and higher selectivity to total oxidation products and hydrocarbon by-products. Before ignition of the gas phase the hydrogen selectivity is increased in OSR as compared to POX, the main contribution being the water-gas shift reaction. After gas phase ignition, increased formation of hydrocarbon by

  1. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  2. Research reactor instrumentation and control technology. Report of a technical committee meeting

    International Nuclear Information System (INIS)

    The majority of research reactors operating today were put into operation 20 years ago, and some of them underwent modifications, upgrading and refurbishing since their construction to meet the requirements for higher neutron fluxes. However, a few of these ageing research reactors are still operating with their original instrumentation and control systems (I and C) which are important for reactor safety to guard against abnormal occurrences and reactor control involving startup, shutdown and power regulation. Worn and obsolete I and C systems cause operational problems as well as difficulties in obtaining replacement parts. In addition, satisfying the stringent safety conditions laid out by the nuclear regulatory bodies requires the modernization of research reactors I and C systems and integration of additional instrumentation units to the reactor. In order to clarify these issues and to provide some guidance to reactor operators on state-of-art technology and future trends for the I and C systems for research reactors, a Technical Committee Meeting on Technology and Trends for Research Reactor Instrumentation and Controls was held in Ljubljana, Slovenia, from 4 to 8 December 1995. This publication summarizes the discussions and recommendations resulting from that meeting. This is expected to benefit the research reactor operators planning I and C improvements. Refs, figs, tabs

  3. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  4. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  5. Investigation of small and modular-sized fast reactor

    International Nuclear Information System (INIS)

    In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

  6. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  7. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  8. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  9. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  10. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  11. Fusion reactor materials

    International Nuclear Information System (INIS)

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  12. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  13. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  14. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  15. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  16. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  17. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  18. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP)

  19. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  20. Fusion reactor materials

    International Nuclear Information System (INIS)

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

  1. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  2. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1980 are described. The work is presented in three chapters: General Information on the Department, Summary of the Department's Development during 1980, and Activities of the Department. Lists of staff, publications, computer programs, and test facilities are included. (author)

  3. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  4. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. The development of a nuclear rocket engine reactor (NRER) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  5. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  6. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  7. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis

  8. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  9. Reactor physics problems on HCPWR

    International Nuclear Information System (INIS)

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  10. Development of Digital MMIS for Research Reactors: Graded Approaches

    International Nuclear Information System (INIS)

    Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

  11. A concept of an advanced inertia fusion reactor; TAKANAWA-I

    International Nuclear Information System (INIS)

    A concept of an advanced inertia fusion reactor: TAKANAWA-I is proposed. A pellet with DT ignitor and DD major fuel, Pb wet walls, C or SiC blocks for shielding, and SiC vessels in the water pool are employed. This reactor does not need blanckets for T breeding, since T is supplied through DD reaction, and has low induced radioactivities. These and a simple structure might give a hopeful prediction of economical and safe advantages and mitigate difficulties of reactor technologies, especially remote maintenance of the reactor. (author)

  12. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  13. Investigation of neutron distribution in training reactor VR-1

    International Nuclear Information System (INIS)

    The VR-1 training reactor is a pool-type light-water reactor with the low-enriched uranium and maximum thermal power of 1 kW. The reactor is mainly used for students' training. The training is aimed to areas such as the reactor physics, neutronics, dosimetry, nuclear safety and I and C systems. Since neutron flux in the VR-1 core is well measured, this work focuses on one part of the reactor - its Radial experimental Channel (RC). This paper deals with the measurement of the neutron distribution by means of gold-foil neutron-activation technique and continual measurement with 3He-filled detector. Obtained experimental results were verified with the simulation in the Monte-Carlo N-Particle Transport Code. Results and conclusions from this paper will be used for further investigation of neutrons and their spatial distribution inside the low-power training reactor. Also, the data obtained in this paper can be used as a basis for future detailed measurements of neutron flux and its distribution in other hard accessible areas inside the reactor. The paper gives a simple theoretical introduction concerning neutron measurement procedures and available techniques in this field, which is particularly important for improving training courses and a content of offered experiments in the VR-1 reactor. (author)

  14. Triga mark-II,III reactor safety re-evaluation

    International Nuclear Information System (INIS)

    In order to revise safety analysis report of old TRIGA reactors, safety re-evaluation of these reactor was started for necessary parts. This report contains the first year results of the project scheduled for two years. The guide lines of safety re-evaluation was made by translating that of nuclear power plant from the view point of TRIGA reactor confirming the basic safety philosophy as much as possible. First of all, sections of reactor history and comparison with similar reactors are made, since the actual operation records, changes, any modification of similar reactors constructed after then, etc., are realistic and valuable data from the safety aspect of old reactor. For the effectiveness of nuclear analysis, a PC based analysis system using WIMS-D/4 and VENTURE was established, and a program for the natural convection cooling analysis of TRIGA reactor was developed. As a result of thermal-hydraulic analysis it was confirmed that the operation limit of fuel temperature set at 650 deg C without any logical reason is very close to the DNB limit. (Author)

  15. Oxidative coupling of methane using inorganic membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Y.H.; Moser, W.R.; Dixon, A.G. [Worcester Polytechnic Institute, MA (United States)] [and others

    1995-12-31

    The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.

  16. Energy analysis and carbon dioxide emission of Tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Energy gain and carbon dioxide (CO2) emission of tokamak fusion power reactors are evaluated in this study compared with other reactor types, structural materials, and other Japanese energy sources currently in use. The reactors treated in this study are (1) a conventional physics performance international thermonuclear experimental reactor (ITER), like a reactor based upon the ITER engineering design activity (ITER-EDA), (2) a RS (reversed shear) reactor using the reversed shear safety-factor/plasma current profile, and (3) a ST (spherical torus) reactor based upon the final version of the advanced reactor innovative engineering study ST (ARIES-ST). The input energy and CO2 emission from these reactors are calculated by multiplying the weight or cost of the fusion reactor components by the energy intensity and/or with the CO2 intensity data, which are updated as often as possible. The ITER cost estimation is estimated based on the component unit costs. The following results were obtained: (1) The RS and the ST reactor can double the energy gain and reduce CO2 emission by one-half compared with the ITER-like reactor. (2) Silicon carbide (SiC) used as the structural material of inner vessel components is best for energy gain and CO2 emission reduction. (3) The ITER-like reactor is slightly superior to a photovoltaic (PV) with regard to CO2 emission. (4) The energy gain and CO2 emission intensity of the RS reactor and the ST reactor are as excellent as those of a fission reactor and a hydro-powered generator. These results indicate that a tokamak fusion power reactor can be one of the most effective power-generating technologies both in high-energy payback gains and reduction of CO2

  17. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  18. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  19. Reactor vessel support system. [LMFBR

    Science.gov (United States)

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  20. Reactor monitoring using antineutrino detectors

    Science.gov (United States)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  1. Reactor physics activities in Japan

    International Nuclear Information System (INIS)

    This report reviews the research activity in reactor physics field in Japan during July, 1992 - July, 1993. The review was performed in the following fields : nuclear data evaluation, calculational method development, fast reactor physics, thermal reactor physics, advanced core design, fusion reactor neutronics, nuclear criticality safety, shielding, incineration of radioactive nuclear wastes, noise analysis and control and national programs. The main references were taken from journals and reports published during this period. The research committee of reactor physics is responsible for the review work. (author)

  2. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  3. Experimental loop in the Nuclear Training Reactor Budapest

    Energy Technology Data Exchange (ETDEWEB)

    Csom, Gy.; Kocsis, E.; Zsolnay, E.M.; Szondi, E.J.; Szuecs, I. (Budapesti Mueszaki Egyetem (Hungary). Egyetemi Reaktor)

    1982-01-01

    The in-pile loop built into the Nuclear Training Reactor of the Technical University Budapest constructed jointly with the specialists of the Moscow Energetic Institute is used for thermoradiolitic investigations of irradiated solutions under conditions of 20-300 deg C temperature and max. 150 bar pressure. Therefore, the results of such experiments can provide valuable information on the kinetics and mechanisms of chemical processes occurring in the primary and secondary circuits of WWER-type power reactors. In order to obtain results applicable also to power plant conditions, the dose rate had to be increased. Therefore, some modifications of the reactor power were necessary. Preliminary test results are summarized.

  4. Research means to back the development of nuclear reactors

    International Nuclear Information System (INIS)

    After 50 year long feedback experience on nuclear reactor operations it is legitimate to wonder whether experimental facilities used to support nuclear power programs are still necessary. The various participants of this conference said yes for mainly 4 reasons: -) to validate the extension of the service life of a reactor without putting at risk its high safety standard, -) to give the reactor more flexibility to cope with the power demand, -) to confront the results given by computerized simulations with experimental data, and -) to qualify the nuclear systems of tomorrow. (A.C.)

  5. Initial prediction of dust production in pebble bed reactors

    Directory of Open Access Journals (Sweden)

    M. Rostamian

    2011-09-01

    Full Text Available This paper describes the computational simulation of contact zones between pebbles in a pebble bed reactor. In this type of reactor, the potential for graphite dust generation from frictional contact of graphite pebbles and the subsequent transport of dust and fission products can cause significant safety issues at very high temperatures around 900 °C in HTRs. The present simulation is an initial attempt to quantify the amount of nuclear grade graphite dust produced within a very high temperature reactor.

  6. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  7. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  8. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  9. 控制黑棒和灰棒对AP1000反应堆Keff值影响的M-C模拟%M-C Simulation on Keff Value for Control and Gray Rod Effect in AP1000 Reactor

    Institute of Scientific and Technical Information of China (English)

    魏强林; 刘义保; 杨波; 吴和喜

    2013-01-01

    采用正在三门建造的AP1000核电厂堆芯参数,使用MCNP5程序建立AP1000堆芯数学模型.考虑了燃料棒、黑棒与灰棒7种不同排布方式,分3种情况通过调节黑棒和灰棒在堆芯中的深度来研究有效增值因数Keff值的变化情况.模拟结果表明:随着黑棒和灰棒在反应堆堆芯中的插入,Keff值在1.44-1.22之间变化.为了验证其合理性,并用1 000 x10-6(ppm)的硼酸溶液进行了化学补偿模拟试验,计算得Keff值在1.17-1.07之间,基本能够满足降低过剩反应性的要求.%Reactor core parameter of API000 nuclear power plant is used. Taking into account 7 different ways arrangement of the fuel rod, control and gray rod, AP1000 reactor core mathematical model is established by the MCNP5 code. The effective multiplication factor Keff is studied through 3 different ways by adjusting the depth of control and gray rod in the reactor core. Simulated results show that the Keff value changed between 1.44 and 1. 22 with insertion of control and gray rod, In order to verify the reasonableness, the simulation experiment of chemical compensation is done by using 1 000 × 10-6 (ppm) boric acid, and the experiments show the Keff values changed between 1.17 and 1.07 , which could basically meet the requirements of reducing the excess reactivity.

  10. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

  11. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIMtm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  12. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  13. Development of a Robust Tri-Carbide Fueled Reactor for Multimegawatt Space Power and Propulsion Applications

    Energy Technology Data Exchange (ETDEWEB)

    Samim Anghaie; Travis W. Knight; Johann Plancher; Reza Gouw

    2004-08-11

    An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors.

  14. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  15. Reconditioning of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    The paper describes the activities carried out to recondition the TRIGA Mark III reactor at the Mexican Nuclear Centre, namely repair of its containment system, maintenance of its operational systems, and the obtaining of a licence for the facility and its operating staff. The process of initially obtaining the operating licence from the regulatory authority was affected by the existence of water leaks in the pool which were detected in March 1985 and were caused by corrosion in the reactor containment system. Reconditioning began with a series of activities aimed at locating, delimiting and repairing the areas damaged by corrosion and involved establishing criteria for selecting the most appropriate inspection, testing and repair methods. In order to obtain the operating licence, it was necessary to comply with various requirements laid down by the regulatory body. The most important requirements included: (a) repair of the reactor pool; (b) maintenance of its operational systems; (c) preparation and implementation of the Quality Control Programme; (d) updating of the Safety Report; (e) updating and preparation of operating, repair, radiation safety, emergency and administrative procedures; and (f) training of operating staff. In addition, the paper describes the work carried out at this reactor to widen its field of research and range of utilization. This work includes the reconditioning of a neutron diffractometer, the design and construction of a neutron diffractometer to determine the textures of materials, and the analysis of a new mixed core configuration based on fuels with 20% and 70% 235U enrichment. (author). 7 refs

  16. The WWR-SM-20 research reactor

    International Nuclear Information System (INIS)

    In this paper the design features and experimental capabilities of the WWR-SM-20 research reactor are described. The reactor uses fuel assemblies consisting of six coaxial fuel tubes with a square cross-section. IRT-3M fuel assemblies can be used with both 90% enriched and 36% enriched uranium. The main characteristics of the IRT-3M fuel assemblies are given, as are the technical and physical parameters of the WWR-SM-20 reactor. The core can hold up to ten ampoule-type channels with a diameter of up to 68 mm. For irradiation purposes, up to 22 26-mm-diameter channels in the fuel assemblies, and up to 48 42-mm-diameter channels in the beryllium blocks of the reflector can be used. In the graphite blanket between the horizontal channels, channels with a diameter of up to 130 mm can be used. The thermal neutron flux density has a maximum value of 1.5 X 1018 m-2 · s-1 in the core and 2.3 X 1018 m-2 · s-1 in the reflector, and the fast neutron flux density (cE > 0.821 MeV) a maximum of 1.9 X 1018 m-2 · s-1. A number of design features have been incorporated in the WWR-SM-20 reactor to make it effectively safe

  17. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  18. Elk River Reactor dismantling

    International Nuclear Information System (INIS)

    The dismantling program was carried out in three overlapping phases: the planning phase which included the preliminary planning and selection of the dismantling approach, the dismantling phase which included all work performed to remove the reactor facility and restore the site to its pre-reactor condition, and the closeout phase which included the final site survey and efforts necessary to terminate the AEC license and contract. Of particular interest was the use of a remotely operated plasma cutting torch to section the pressure vessel internals, the pressure vessel and the outer thermal shield, the use of explosives in removal of the biological shield and the method of establishment of the criteria for material disposal

  19. REACTOR MODERATOR STRUCTURE

    Science.gov (United States)

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  20. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  1. Decommissioning of research reactors

    International Nuclear Information System (INIS)

    Research reactors of WWR-S type were built in countries under Soviet influence in '60, last century and consequently reached their service life. Decommissioning implies removal of all radioactive components, processing, conditioning and final disposal in full safety of all sources on site of radiological pollution. The WWR-S reactor at Bucuresti-Magurele was put into function in 1957 and operated until 1997 when it was stopped and put into conservation in view of decommissioning. Presented are three decommissioning variants: 1. Reactor shut-down for a long period (30-50 years) what would entail a substantial decrease of contamination with lower costs in dismantling, mechanical, chemical and physical processing followed by final disposal of the radioactive wastes. The drawback of this solution is the life prolongation of a non-productive nuclear unit requiring funds for personnel, control, maintenance, etc; 2. Decommissioning in a single stage what implies large funds for a immediate investment; 3. Extending the operation on a series of stages rather phased in time to allow a more convenient flow of funds and also to gather technical solutions, better than the present ones. This latter option seems to be optimal for the case of the WWR-S Research at Bucharest-Magurele Reactor. Equipment and technologies should be developed in order to ensure the technical background of the first operations of decommissioning: equipment for scarification, dismantling, dismemberment in a highly radioactive environment; cutting-to-pieces and disassembling technologies; decontamination modern technologies. Concomitantly, nuclear safety and quality assurance regulations and programmes, specific to decommissioning projects should be implemented, as well as a modern, coherent and reliable system of data acquisition, recording and storing. Also the impact of decommissioning must be thoroughly evaluated. The national team of specialists will be assisted by IAEA experts to ensure the

  2. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  3. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  4. Nuclear reactor container

    International Nuclear Information System (INIS)

    Upon reactor accident, hydrogen and oxygen are generated by water-zirconium reaction and radiolysis of water, which are accumulated in the reactor. If the concentration of hydrogen and oxygen exceeds a burning limit, there is a possibility of hydrogen burning to cause a danger of deteriorating the integrity of the reactor container and the equipments therein. The limit for the occurrence of the detonation is determined by a relationship between the scale of a detonation cell and the size of the container, and if the scale is greater than the container, the detonation does not occur. The scale of the cell is determined by a gas combustion rate and, if the combustion reaction is suppressed, detonation does not occur even in a large container. Then, an appropriate diluent is added to increase heat capacity of a gas mixture to thereby suppress the temperature elevation of the gas. Incombustible gases having a great heat capacity are preferred for the diluent, and CO2 is used. As the concentration of the CO2 gas to be added is increased, the detonation cell is made greater. Thus, occurrence of detonation due to combustion of the accumulated hydrogen can be prevented. (N.H.)

  5. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    In the middle of 21st century, the population on the earth is expected to double, and the energy that mankind consumes to triple. The nuclear fusion which is said the ultimate energy source for mankind is expected to solve this energy problem. As for fusion reactors, fuel materials exist inexhaustibly, distributing evenly, they have high safety in principle, the product of burning is harmless nonradioactive substance that does not require the treatment and disposal, and the attenuation of induced radioactivity due to neutrons is quick and the effect to global environment is little. The basic plan of second stage nuclear fusion research and development was decided in 1975, aiming at attaining the critical plasma condition. JT-60 has attained it in 1987. The project of international thermonuclear fusion experimental reactor (ITER) was started, and the conceptual design was carried out. Under such background, the third stage basic plan was decided in 1992, and its objective is self ignition condition, long time burning and the basis of the reactor engineering technology. The engineering design of the ITER is investigated. (K.I.)

  6. High temperature gas reactor

    International Nuclear Information System (INIS)

    The present invention provides a reflector block structure of a high temperature gas reactor in which graphite blocks are not failed even a containing cylinder loaded to a fuel exchanger collides against to secured reflectors upon loading and withdrawing fuel constitutional elements. Namely, a protection plate made of a metal material such as stainless steel is covered on the secured reflector blocks disposed to the upper most step among secured graphite reflector blocks constituting the reactor core. In addition, positioning guide grooves are formed on the protection plate for guiding the containing cylinder loaded to the fuel exchanger to the column of the reactor core constitutional elements. With such a constitution, even if the containing cylinder of fuel exchanger is hoisted down and collided against the inner circumferential edge of the secured reflector blocks due to deviation of the position and the direction upon exchange of fuels, the reflector blocks are not failed since the above-mentioned portion is covered with the metal protection plate. In addition, the positioning guide grooves lead the fuel exchanger to a predetermined column correctly. (I.S.)

  7. BWR type reactor

    International Nuclear Information System (INIS)

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  8. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report 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 of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  9. Fusion reactor safety

    International Nuclear Information System (INIS)

    Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

  10. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report 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 of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units are provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  11. OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

  12. Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

    International Nuclear Information System (INIS)

    This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes

  13. Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Appel and J. M. Capron

    2007-07-25

    This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

  14. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  15. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  16. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The purpose of this technical paper is to provide status of the United State domestic Research Reactor Infrastructure (RRI) Program at the Idaho National Laboratory. This paper states the purpose of the program, lists the universities operating TRIGA reactors that are supported by the program, identifies anticipated fresh fuel needs for the reactor facilities, discusses spent fuel activities associated with the program, and addresses successes and planned activities for the program. (author)

  17. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic

  18. Applications of Research Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  19. Effect of post-digestion temperature on serial CSTR biogas reactor performance

    DEFF Research Database (Denmark)

    Boe, Kanokwan; Karakashev, Dimitar Borisov; Trably, Eric;

    2009-01-01

    The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT...... of 5.3 days. Three post-digestion temperatures (55 degrees C, 37 degrees C and 15 degrees C) were compared in terms of biogas production, process stability, microbial community and methanogenic activity, The results showed that the post-digesters operated at 55 degrees C, 37 degrees C and 15 degrees C...... gave extra biogas production of 11.7%, 8.4% and 1.2%, respectively. The post-digester operated at 55 degrees C had the highest biogas production and was the most stable in terms of low VFA concentrations. The specific methanogenic activity tests revealed that the main reactor and the post...

  20. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  1. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices

  2. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    Paper dwells upon the design and the operation of one of the German test reactors, namely, the TRIGA Mainz one (TRIGA: Training Research Isotope Production General Atomic). The TRIGA reactor is a pool test reactor the core of which contains a graphite reflector and is placed into 2 m diameter and 6.25 m height aluminum vessel. There are 75 fuel elements in the reactor core, and any of them contains about 36 g of 235U. The TRIGA reactors under the stable operation enjoy wide application to ensure tests and irradiation, namely: neutron activation analysis, radioisotope production, application of a neutron beam to ensure the physical, the chemical and the medical research efforts. Paper presents the reactor basic experimental program lines

  3. Mechanical characterization of superalloys for space reactors

    International Nuclear Information System (INIS)

    The aim of this work is the selection of structural materials that can be used in the temperature range 600-9000C for a gas cooled space reactor producing electricity. Superalloys fit best the temperature range required. Five nickel base alloys are chosen for their good mechanical behaviour: HAYNES 230, HASTELLOY S, HASTELLOY X, HASTELLOY XR and PYRAD 38D. Metallography, tensile and hardness tests are realized. Sample contraction is evidenced for some creep tests, under low stress: 20MPa at 8000C, on HAYNES 230 and HASTELLOY X, probably related to the structural evolution of these materials corresponding to a decrease of the crystal parameter

  4. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  5. Industrializing the liquid metal reactor

    International Nuclear Information System (INIS)

    Commercial acceptance of the liquid metal reactor had its beginning with the Fermi reactor, over two decades ago. The pattern of industrialization since that time is discussed, contrasting domestic and foreign experience. The recent termination of the Clinch River reactor project marks a watershed in the U.S. approach towards commercialization. The increased emphasis on achieving cost competitive designs reflects an awareness that barriers to industrialization are institutional and financial, and not technological

  6. Acceptability of reactors in space

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  7. Hydrogen Production in Fusion Reactors

    OpenAIRE

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

    1993-01-01

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

  8. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  9. Turning points in reactor design

    International Nuclear Information System (INIS)

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

  10. Optimal control of nuclear reactors

    International Nuclear Information System (INIS)

    The modern control theory is applied to the design of control systems for experimental nuclear reactors that do not belong to power reactors, the component forms of optimal control systems for nuclear reactors are demonstrated. The adoption of output quadratic integral criterion and incomplete state feedback technique can make these systems both efficient and economical. Moreover, approximate handling methods are given so as to simplify the calculations in design. In addition, the adoptable reference values of parameters are given in the illustration

  11. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  12. Acceptability of reactors in space

    International Nuclear Information System (INIS)

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System

  13. Advanced Fission Reactor Program objectives

    International Nuclear Information System (INIS)

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  14. Reactor safety - an international task

    International Nuclear Information System (INIS)

    The dimensions and the significance of the task of ensuring reactor safety can be defined on the basis of experiences gained from Harrisburg and Chernobyl. The countries that use nuclear energy are tied together to a community by virtue of the risk they share. Therefore the GRS is working in close cooperation with the EC, OECD, IAEO and COMECON. This results in safety examinations of the Greifswald reactor, safety analyses of nuclear reactors in Germany, France and the USA and also considerations on the safety demands to be placed on new reactor concepts. (DG)

  15. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  16. Integrated modular water reactor: IMR

    International Nuclear Information System (INIS)

    The Mitsubishi Heavy Industries, Ltd. Has investigated on a concept on small scale reactor with economical efficiency comparable with large scale one. Aims of development on the integrated modular water reactor (IMR) of a small scale reactor plant concept consist in large construction cost reduction through adoption of technique specific to the small scale reactor and integrated production of plural units and in establishment of high safety target without reality in a large scale reactor to realize reduction of operation and maintenance costs by this reduction to simplification of operation and maintenance. Its concrete developmental targets are to make an integrated reactor with vessel size actually producible and the largest output, to remove feasibility of coolant loss accident (LOCA), to remove an accident with feasibility related to fuel fracture, to remove feasibility of nuclear reactor coolant to leak out from a storage vessel, to secure safety of plant without necessity of human and physical assistances from other plants at all on an accident, to make numbers of operators per unit output equal to those of large scale reactor, and to make working amounts at maintenance per unit output equal to large scale reactor by simplification of apparatus practice of rotation on main apparatus such as SG, and so on. Here were described on design concept and plan to realization. (G.K.)

  17. 3. Interindustry conference on reactor materials science

    International Nuclear Information System (INIS)

    This document contains abstracts on papers presented at the Third Interindustry Conference on Reactor Materials Science (Dimitrovgrad, 27-30 October 1992). The subject scope of the papers is a follows: fuel and fuel elements of power reactors; structural materials of fast breeder reactors and thermonuclear reactors; structural materials of WWER and RBMK type reactors; absorbers and moderators

  18. Helias reactor studies

    International Nuclear Information System (INIS)

    The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. The magnetic field has 5 field periods and the main optimization principle is the reduction of the Pfirsch-Schlueter currents and the Shafranov shift, which has been verified by computations with the NEMEC and MFBE-codes. The modular coil system comprises 50 coils, which are constructed using NbTi-superconducting cables. The basic dimensions are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 5 T, maximum field on the coils 10 T. Forces and stresses in the coil system have been investigated with the aid of the ANSYS code, which found maximum stress values of about 650 MPa in the coil casing. Helias configurations with 4 and 3 field periods have been constructed by starting from the 5-period case and by eliminating one or two periods while the shape of the coils is kept nearly invariant. In a first survey blanket concepts, developed for the DEMO tokamak, have been adapted to the Helias geometry, in particular, the solid breeder concept developed by FZK (Karlsruhe) has been extrapolated to the Helias geometry identifying the drawbacks and advantages of this concept. Furthermore, the liquid breeder concept using Li7-Pb83 and water-cooling is an interesting alternative for the Helias reactor. Maintenance of blanket and plasma facing components is possible through the portholes between modular coils. Numerical simulations of the start-up phase of the Helias reactor using the TOTAL-P code have confirmed the zero-dimensional modeling of the fusion plasma with the aid of empirical scaling laws. (author)

  19. Photocatalytic reactors for treating water pollution with solar illumination: a simplified analysis for n-steps flow reactors with recirculation

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Universitaet Hannover (Germany). Institut fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [INTEC Universidad Nacional del Litoral and CONICET, Sante Fe (Argentina)

    2005-09-01

    The concentration of dissolved oxygen in water, in equilibrium with atmospheric air (ca. 8 ppm at 20{sup o}C), defines the limits of all practical oxidizing processes for removing pollutants in photocatalytic reactors. To solve this limitation, an alternative approach to that of a continuously aerated reactor is the use of a recirculating system with aeration performed after every cycle at the reactor entering stream. As defined by the nature of a single recirculating step (the need of a reactor operation at a rather low concentration range), this procedure results in a very low photonic efficiency (thus requiring a large photon collecting area and consequently increasing the capital cost). The design engineer will have to resort to a series of several reactors with recirculation. This solution may then lead to a very high Photonic Efficiency for the entire process (i.e., a reduced light harvesting area) at the price of an increase in the required capital cost (due to the larger number of reactors). This paper provides a very simple analysis and analytical expressions that can be used to estimate, for a desired degree of degradation, a trade-off solution between a high number of reactors and a very large surface area to collect the solar photons. (author)

  20. Intensification of Deep Hydrodesulfurization Through a Two-stage Combination of Monolith and Trickle Bed Reactors

    Institute of Scientific and Technical Information of China (English)

    Min Xu; Hui Liu⁎; Shengfu Ji; Chengyue Li

    2014-01-01

    Deep hydrodesulfurization (HDS) is an important process to produce high quality liquid fuels with ultra-low sul-fur. Process intensification for deep HDS could be implemented by developing new active catalysts and/or new types of reactors. In this work, the kinetics of dibenzothiophene (DBT) hydrodesulfurization over Ni-P/SBA-15/cordierite catalyst was investigated at 340-380 °C and 3.0-5.0 MPa. The first-order reaction model with respect to both DBT and H2 was used to fit the kinetics data in a batch recycle operation system. It is found that both the activation energy and rate constant over the Ni-P monolithic catalyst under our operating conditions are close to those over conventionally used HDS catalysts. Comparative performance studies of two types of reactors, i.e., trickle bed reactor and monolithic reactor, were performed based on reactor modeling and simulation. The results indicate that the productivity of the monolithic reactor is 3 times higher than that of the trickle bed reactor on a catalyst weight basis since effective utilization of the catalyst is higher in the monolithic reactor, but the volumetric productivity of the monolithic reactor is lower for HDS of DBT. Based on simulation results, a two-reactor-in-series configuration for hydrodesulfurization is proposed, in which a monolithic reactor is followed by a tickled bed reactor so as to attain intensified performance of the system converting fuel oil of different sulfur-containing compounds. It is il ustrated that the two reactor scheme outperforms the trickle bed reactor both on reactor volume and catalyst mass bases while the content of sulfur is reduced from 200μg·g-1 to about 10μg·g-1.

  1. Reactor coolant pump flywheel

    Science.gov (United States)

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  2. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report 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 of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page

  3. FUEL ASSAY REACTOR

    Science.gov (United States)

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  4. Shutting down two reactors

    International Nuclear Information System (INIS)

    Nuclear power will be phased out of the swedish energy system during the first decades of the next century. Commissioned by the swedish government, the National Energy Administration reports a study on the possibilities for, and consequences of, an earlier shut down (1994-1996) of two of the twelve swedish power reactors. Some of the questions studied are: How much will the electricity price raise ?; How will the electricity consumption be affected ?; What are the alternatives to nuclear power ?; What will the cost be ? and What will the environmental effects be ?. (L.E.)

  5. Measurement in nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor construction has a flux detector comprising a bundle of fibre optics each having a bead incorporating a substance which scintillates on being struck by neutrons or gamma radiations. The other ends of the fibre optics terminate at an image intensifier. The optical fibres may be of glass made from a mixture of silica, alkaline earth metal oxide, cerous oxide and alkali metal oxide. The beads may be incorporated in a disc forming a detector head, which is in a protective guide tube, through which an inert gas may be passed. (author)

  6. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  7. Reactor system safety assurance

    International Nuclear Information System (INIS)

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  8. The Oklo reactors

    International Nuclear Information System (INIS)

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

  9. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  10. Reactor surface contamination stabilization. Innovative technology summary report

    International Nuclear Information System (INIS)

    Contaminated surfaces, such as the face of a nuclear reactor, need to be stabilized (fixed) to avoid airborne contamination during decontamination and decommissioning activities, and to prepare for interim safe storage. The traditional (baseline) method of fixing the contamination has been to spray a coating on the surfaces, but ensuring complete coverage over complex shapes, such as nozzles and hoses, is difficult. The Hanford Site C Reactor Technology Demonstration Group demonstrated innovative technologies to assess stabilization properties of various coatings and to achieve complete coverage of complex surfaces on the reactor face. This demonstration was conducted in two phases: the first phase consisted of a series of laboratory assessments of various stabilization coatings on metal coupons. For the second phase, coatings that passed the laboratory tests were applied to the front face of the C Reactor and evaluated. The baseline coating (Rust-Oleum No. 769) and one of the innovative technologies did not completely cover nozzle assemblies on the reactor face, the most critical of the second-phase evaluation criteria. However, one of the innovative coating systems, consisting of a base layer of foam covered by an outer layer of a polymeric film, was successful. The baseline technology would cost approximately 33% as much as the innovative technology cost of $64,000 to stabilize an entire reactor face (196 m2 or 2116 ft2) with 2,004 nozzle assemblies, but the baseline system failed to provide complete surface coverage

  11. A design algorithm for batch stirred tank transesterification reactors

    International Nuclear Information System (INIS)

    Highlights: ► Simplified algorithm for batch biodiesel reactor design was developed. ► C Sharp software tool for implementing the algorithm was also developed. ► 50 L/batch reactor was constructed and used to process neem oil biodiesel. ► Results showed that the produced neem biodiesel is a fuel grade product. ► Scale-up of the reactor was carried out using the developed software. - Abstract: A 50 L per batch, stirred tank reactor, suitable for carrying out transesterification of vegetable oils was designed and constructed. The major design assumptions included stainless steel plate thickness of 2 mm, reaction temperature of 60–65 °C and an initial/final fluid temperature of 25/70 °C. The calculated impeller Reynolds number was in the mixed regime zone of 10–104; the power number was varied between 1 and 5, while a typical propeller speed of 22.5 rev/s (or 1350 rev/min) was adopted. The limiting design conditions were maximum reactor diameter of 1.80 m, straight side height-to-diameter ratio in the range of 0.75–1.5 and minimum agitator motor power of 746 W (1 Hp). Based upon the design, a simple algorithm was developed and interpreted into Microsoft C Sharp computer programming language to enable scale up of the reactor. Performance testing of the realized reactor was carried out while using it to produce Neem oil biodiesel via base – catalyzed methanolysis, which yielded high quality fuel product.

  12. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  13. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  14. Analysis and upgrade of instrumentation and control systems for the modernization of research reactors

    International Nuclear Information System (INIS)

    This document provides assistance in the review and planning process for the upgrade of instrumentation and control systems (I and C systems) and related safety features of the reactor protection system for research reactors. In the interest of safety a need was realized to evaluate the performance of outdated I and C systems. An advisory group was assembled to develop guidelines and to provide recommendations for the upgrade of I and C systems. The recommendations on I and C systems upgrade contained in this document were developed by the advisory group using as guidelines the established safety criteria and operating standards for research reactors. 24 refs

  15. OXIDATIVE COUPLING OF METHANE USING INORGANIC MEMBRANE REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Y.H. Ma; Dr. W.R. Moser; Dr. A.G. Dixon; Dr. A.M. Ramachandra; Dr. Y. Lu; C. Binkerd

    1998-04-01

    The objective of this research is to study the oxidative coupling of methane in catalytic inorganic membrane reactors. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and higher yields than in conventional non-porous, co-feed, fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for the formation of CO{sub x} products. Such gas phase reactions are a cause of decreased selectivity in the oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Membrane reactor technology also offers the potential for modifying the membranes both to improve catalytic properties as well as to regulate the rate of the permeation/diffusion of reactants through the membrane to minimize by-product generation. Other benefits also exist with membrane reactors, such as the mitigation of thermal hot-spots for highly exothermic reactions such as the oxidative coupling of methane. The application of catalytically active inorganic membranes has potential for drastically increasing the yield of reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity.

  16. Methane coupling by membrane reactor. Quarterly technical progress report, September 25, 1995--December 24, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-15

    The performance of the third type of catalytic membrane reactor configuration, with catalyst deposited in the membrane and no catalyst or inert materials in the tube side, was evaluated. The C{sub 2} selectivity obtained was about 10% due to the gas phase reaction in the empty tube side of the reactor. The membrane reactor with an oxygen-permeable dense membrane has been built. The use of a dense membrane will eliminate the loss of hydrocarbon from the tube side to the shell side, as observed in the Vycor glass membrane reactor. Also, air can be used as the oxygen source without contaminating the product. La/MgO was synthesized and will be used as the catalyst for the dense membrane reactor. This catalyst was reported in the literature to show significant improvement of C{sub 2} selectivity and yield for oxidative coupling of methane in a packed-bed reactor by using the operation mode of staged-feed of oxygen. A reactor mode for methane oxidative coupling in reactors with both distributed oxygen feed and C{sub 2} product removal was developed based on the general model of cross-flow reactors reported in the last quarterly report. A distributed oxygen feed could give rise to much higher C{sub 2} yield than the co-feed reactor as long as the space time is long enough. In the case of a two-membrane reactor, where oxygen is supplied by one membrane and products are removed through the other membrane, a high separation factor of C{sub 2} product to methane for the product-removal membrane is critical to achieve high C{sub 2} yield.

  17. Fischer-Tropsch Slurry Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Y.; Gamwo, I.K.; Harke, F.W. [Pittsburgh Energy Technology Center, PA (United States)] [and others

    1995-12-31

    This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas, solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.

  18. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  19. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  20. Reactor system on barge

    International Nuclear Information System (INIS)

    Floating electrical power plants or power plant barges add new dimensions to utility planners and agencies in the world. Intrinsically safe and economical reactors (ISER) employ steel reactor pressure vessels, which significantly reduce the weight as compared with PIUS, and provide siting versatility including barge-mounted plants. In this paper, the outline of power plant barges and barge-mounted ISERs is described. Besides their mobility, power plant barges have the salient advantages such as short delivery time and better quality control due to the outfitting in shipyards. These power plant barges may be temporarily moored or permanently grounded in shallow water at the centers of industrial complexes or the suitable areas adjacent to them, and satisfy the increasing needs for electric power. A cost-effective and technically perfect barge positioning system should be designed to meet the specific requirement for the location and its condition. Offshore siting away from coast may be applicable only to large plants of 1,000 MWe or more, and inshore siting and coastal or river siting are considered for an ISER-200 barge-mounted plant. The system of a barge-mounted ISER plant is discussed in the case of a floating type and the type on a seismic base isolator. (Kako, I.)

  1. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  2. The EPR reactor NDE

    International Nuclear Information System (INIS)

    In May 2006, Electricite de France decided to launch the building of the first EPRR Reactor on the Flamanville site in Normandy. The 'Flamanville 3' EPR unit is the first one to be subjected to the French Ministerial Orders of the 10 November 1999 and of the 13 December 2005 from the design phase. According to these orders, the non destructive examination (NDE) planned for the in service inspection (ISI) and for the pre service inspection (PSI) must be operational with a compulsory formal qualification. The PSI is a complete inspection of the main primary and secondary systems. The PSI's objective is to perform before the first core loading all the NDE planned for the future ISI in the same conditions, in order to have a reliable reference for the detection or for the evaluation of the possible damages during the ISI. The 'Flamanville 3' PSI is planned to start end 2010. The program consists of the development and the qualification of the NDE compatible with this new generation reactor's challenges. The paper is about: - the main EPR's objectives and the technological evolutions, - the main component modifications (which have an impact on the NDE), - the place of ISI in the general safety demonstration, - the main inspection objectives, - the NDE qualification process, - the approach to set up the ISI program, - the ISI program. (authors)

  3. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  4. Status of Japanese university reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Yoshiaki [Research Reactor Institute, Kyoto Univ., Kumatori, Osaka (Japan)

    1999-08-01

    Status of Japanese university reactors, their role and value in research and education, and the spent fuel problem are presented. Some of the reactors are now faced by severe difficulties in continuing their operation services. The point of measures to solve the difficulties is suggested. (author)

  5. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  6. Reactor Neutrino Physics -- An Update

    OpenAIRE

    Boehm, Felix

    1999-01-01

    We review the status and the results of reactor neutrino experiments. Long baseline oscillation experiments at Palo Verde and Chooz have provided limits for the oscillation parameters while the recently proposed Kamland experiment at a baseline of more than 100km is now in the planning stage. We also describe the status of neutrino magnetic moment experiments at reactors.

  7. Chemical-vapor-deposition reactor

    Science.gov (United States)

    Chern, S.

    1979-01-01

    Reactor utilizes multiple stacked trays compactly arranged in paths of horizontally channeled reactant gas streams. Design allows faster and more efficient deposits of film on substrates, and reduces gas and energy consumption. Lack of dead spots that trap reactive gases reduces reactor purge time.

  8. Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

    2006-02-15

    The primary components such as a reactor core, heat exchangers, pumps and internal structures of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A{sub 0} plate wave was selected as the application mode of the sensor. The A{sub 0} plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A{sub 0} mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote

  9. Cooling system for reactor container

    International Nuclear Information System (INIS)

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  10. Engineering reactors for catalytic reactions

    Indian Academy of Sciences (India)

    Vivek V Ranade

    2014-03-01

    Catalytic reactions are ubiquitous in chemical and allied industries. A homogeneous or heterogeneous catalyst which provides an alternative route of reaction with lower activation energy and better control on selectivity can make substantial impact on process viability and economics. Extensive studies have been conducted to establish sound basis for design and engineering of reactors for practising such catalytic reactions and for realizing improvements in reactor performance. In this article, application of recent (and not so recent) developments in engineering reactors for catalytic reactions is discussed. Some examples where performance enhancement was realized by catalyst design, appropriate choice of reactor, better injection and dispersion strategies and recent advances in process intensification/ multifunctional reactors are discussed to illustrate the approach.

  11. Directions in advanced reactor technology

    International Nuclear Information System (INIS)

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  12. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  13. Antineutrino Monitoring of Thorium Reactors

    CERN Document Server

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  14. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  15. Automated reactor records evaluation framework

    International Nuclear Information System (INIS)

    The only truly reliable method for core physics code validation is comparison against experimental data – and for power nuclear reactors, the only reasonably acquirable kind of experimental data are the reactor records. However, the amount of the data coming from the reactor operation is often so vast that it can be discouraging for the code developers to use it properly. Thus, the validation package is further reduced because the data is hard to use. This paper presents an elaborate, fully automated framework, which was designed and implemented in our institute, for reactor records processing and its use for core physics code validation. The workflow, implemented as a Web 2.0 application, provides a practical and painless solution for use of reactor records data for code development and validation. (author)

  16. Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

    Science.gov (United States)

    Elmitwalli, Tarek; Otterpohl, Ralf

    2011-01-01

    The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C.

  17. Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

    Science.gov (United States)

    Elmitwalli, Tarek; Otterpohl, Ralf

    2011-01-01

    The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C. PMID:22097038

  18. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

  19. Neutrino Experiments at Reactors

    Science.gov (United States)

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  20. Licensed operating reactors

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar 1990) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

  1. Reactor safety systems

    International Nuclear Information System (INIS)

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.)

  2. Reactor control rod

    International Nuclear Information System (INIS)

    Object: To enable quick descent of a control rod body even when some relative phase deviation between upper drive means and wrapper tube is produced, while permitting a coolant to effectively flow into a protective tube irrespective of the position of the control rod body. Structure: In a control rod used for a nuclear reactor such as a fast breeder, an orifice which dispenses with a cylindrical guide tube and has a greater inner diameter than the outer diameter of the protective tube of the control rod body is provided on the inner side of a wrapper tube, thus permitting smooth operation of the control rod body and also permitting the coolant to effectively flow into the protective tube irrespective of the control rod body. (Horiuchi, T.)

  3. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  4. Nuclear reactor measurement system

    International Nuclear Information System (INIS)

    An instrument to detect the temperature and flow-rate of the liquid metal current of a coolant fluid sample from adjacent sub-assemblies of a liquid metal-cooled nuclear reactor is described. It includes three thermocouple hot junctions mounted in series, each intended for exposure to a sample-current from a single sub-assembly, electromagnetic coils being mounted around an induction core which detects variations in the liquid metal flow-rate by deformation of the lines of flux. The instrument may also include a thermocouple to detect the mean temperature of the sample-current of coolant fluid from several sources, the result being that the temperature of the coolant fluid current in a sub-assembly may be inferred from the three temperature readings associated with this sub-assembly

  5. OECD: Halden reactor project

    International Nuclear Information System (INIS)

    The work at the Project has continued in the two main fields: test fuel irradiation and fuel research, and computer based process supervision and control. Organizations participating in the Project continue to have their fuel irradiated in the Halden Reactor in instrumented test assemblies designed and manufactured by the Project. The Project's fuel studies continue to focus on specific subjects such as fuel pellet/cladding interaction and heat transfer, fission product release and fuel behavior under loss of coolant conditions. The work on process control and supervision continues in the highly relevant fields of core control and operator-process communication. A system for predictive core control is being developed while special mathematical methods for core power distribution control are being studied. Operator-process communication studies comprise use of computer simulation on colour display as important ingredients, while the work on developing a system for interactive plant disturbance analysis continues

  6. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  7. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  8. Solid oxide electrochemical reactor science.

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)

    2010-09-01

    Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.

  9. Power Reactors. Appendix VIII

    International Nuclear Information System (INIS)

    Decommissioning of nuclear facilities in many countries has evolved into a mature industry that has benefited from experience gained from previous projects and decommissioning costs can now be estimated to a good degree of accuracy. As a result of lessons learned, future decommissioning projects can be performed with higher levels of efficiency. Decommissioning of old power reactors is in progress in several countries. In some cases, decommissioning has been completed (i.e. plant sites have been released from regulatory control), while in other countries decommissioning is still in progress. Several large power reactors have been successfully decommissioned since 1995. The key areas of particular importance for decommissioning are decontamination, radiation protection, dismantling and demolition. The technologies which can be used for these tasks are commonly available on the market, but effective decommissioning still depends on an optimal choice of technologies, including site specific developments. It is not possible to recommend the use of a single specific technology for dismantling, demolition, segmentation or decontamination; rather, it is good practice to take into account as much information as possible from other decommissioning projects and to draw comparisons between various techniques in order to choose the one with the best performance in a particular situation. The exchange of information on all types of decommissioning experience, including decommissioning techniques and their applicability as well as disadvantages for specific tasks, is taking place on various levels, such as: — Collaborative working groups established by international organizations such as the IAEA, the OECD Nuclear Energy Agency and the European Commission and the publication of technical reports by such organizations; — National and international conferences; — Bilateral or multilateral cooperation and information exchange between organizations with responsibilities for

  10. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  11. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Following the accident at Chernobyl nuclear reactor, WHO organized on 6 May 1986 in Copenhagen a one day consultation of experts with knowledge in the fields of meteorology, radiation protection, biological effects, reactor technology, emergency procedures, public health and psychology in order to analyse the development of events and their consequences and to provide guidance as to the needs for immediate public health action. The present report provides detailed information on the transportation and dispersion of the radioactive material in the atmosphere, especially volatile elements, during the release period 26 April - 5 May. Presented are the calculated directions and locations of the radioactive plume over Europe in the first 5 days after the accident, submitted by the Swedish Meteorological and Hydrological Institute. The calculations have been made for two heights, 1500m and 750m and the plume directions are grouped into five periods, covering five European areas. The consequences of the accident inside the USSR and the radiological consequences outside the USSR are presented including the exposure routes and the biological effects, paying particular attention to iodine-131 effects. Summarized are the first reported measured exposure rates above background, iodine-131 deposition and concentrations in milk and the remedial actions taken in various European countries. Concerning the cesium-137 problem, based on the UNSCEAR assessment of the consequences of the nuclear fallout, one concludes that the cesium contamination outside the USSR is not likely to cause any serious problems. Finally, the conclusions and the recommendations of the meeting, taking into account both the short-term and longer term considerations are presented

  12. Nuclear reactor fuelling machine

    International Nuclear Information System (INIS)

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 1800 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 1800 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  13. Reliability of reactor materials

    International Nuclear Information System (INIS)

    This report is the final technical report of the fracture mechanics part of the Reliability of Reactor Materials Programme, which was carried out at the Technical Research Centre of Finland (VTT) through the years 1981 to 1983. Research and development work was carried out in five major areas, viz. statistical treatment and modelling of cleavage fracture, crack arrest, ductile fracture, instrumented impact testing as well as comparison of numerical and experimental elastic-plastic fracture mechanics. In the area of cleavage fracture the critical variables affecting the fracture of steels are considered in the frames of a statistical model, so called WST-model. Comparison of fracture toughness values predicted by the model and corresponding experimental values shows excellent agreement for a variety of microstructures. different posibilities for using the model are discussed. The development work in the area of crack arrest testing was concentrated in the crack starter properties, test arrangement and computer control. A computerized elastic-plastic fracture testing method with a variety of test specimen geometries in a large temperature range was developed for a routine stage. Ductile fracture characteristics of reactor pressure vessel steel A533B and comparable weld material are given. The features of a new, patented instrumented impact tester are described. Experimental and theoretical comparisons between the new and conventional testers indicated clearly the improvements achieved with the new tester. A comparison of numerical and experimental elastic-plastic fracture mechanics capabilities at VTT was carried out. The comparison consisted of two-dimensional linear elastic as well as elastic-plastic finite element analysis of four specimen geometries and equivalent experimental tests. (author)

  14. Carbon dioxide direct cycle modular reactor

    International Nuclear Information System (INIS)

    Recently, as the micro gas-turbine power generation is clean for environment and has high convenience, it is focused as a small size dispersion electric source for super markets, hospitals, factories, and so on. And, a modular high temperature gas reactor (PBMR) adopting the gas turbine is also focused recently, and is progressed on its construction in South Africa and reported on construction plan of the Exelon Inc. in U.S.A. PBMR has specific safety for a small size and pebble-bed reactor and also has some characters on low construction cost similar to that of LWR due to simplification and small size module adoption of its plant. The PBMR uses helium for its coolants, of which exit temperature is set for at 900degC to get higher thermal efficiency. This is because of its adoption of Brayton cycle to fast reduce the efficiency with falling temperature. However, as helium is a costly and easy-emission vapor, it is desired to alternate to cheaper and more difficult-emission vapor. Here were introduced on carbon dioxide (CO2) direct cycle using carbon dioxide with extremely higher thermal efficiency than helium and its applicability to nuclear reactors. (G.K.)

  15. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  16. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

  17. Catalytic Pyrolysis of Oak via Pyroprobe and Bench Scale, Packed Bed Pyrolysis Reactors

    Science.gov (United States)

    The pyrolytic conversion of oak sawdust at 500°C in flowing He over eight proprietary catalysts is described and compared to the control bed material, quartz sand. The reactions were conducted and compared in two reactors, an analytical, ug-scale pyroprobe reactor and a bench, g-scale packed bed re...

  18. 75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Science.gov (United States)

    2010-09-17

    ... which authorizes the possession, use, and operation of the Aerotest Radiography and Research Reactor... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order... Regulations (10 CFR) Section 50.21(c) for research and development purposes. Aerotest is a wholly...

  19. On the development of fast breeder reactors and the use of thorium in Brazil

    International Nuclear Information System (INIS)

    This work presents a discussion on the possibility of construction of fast breeder reactors in Brazil. It is specially concerned with the use of thorium which is abundant in our country. The main advantages of this projects are: develop fuel and reactor technology in Brazil, increase thorium research, demonstrate the safety of LMFBR and promote its public acceptance. (A.C.A.S.)

  20. A review of the United Kingdom fast reactor programme - March 1984

    International Nuclear Information System (INIS)

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (a) Full power operation of the PFR. (b) Development work supporting the NNC designed CDFR. (c) Demonstration and development of the fast reactor fuel cycle