WorldWideScience

Sample records for reactor performance characteristics

  1. Research about reactor operator's personality characteristics and performance

    International Nuclear Information System (INIS)

    Wei Li; He Xuhong; Zhao Bingquan

    2003-01-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  2. Research about reactor operator's personability characteristics and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

    2003-03-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  3. Description and performance characteristics for the neutron Coincidence Collar for the verification of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1981-08-01

    An active neutron interrogation method has been developed for the measurement of 235 U content in fresh fuel assemblies. The neutron Coincidence Collar uses neutron interrogation with an AmLi neutron source and coincidence counting the induced fission reaction neutrons from the 235 U. This manual describes the system components, operation, and performance characteristics. Applications of the Coincidence Collar to PWR and BWR types of reactor fuel assemblies are described

  4. A parametric study on characteristics for nuclear design of high-performance research reactor

    International Nuclear Information System (INIS)

    Joe, D. G.; Lee, C. S.; Lee, B. C.; Seo, C. G.; Chae, H. T.; Park, C.

    2003-01-01

    A conceptual design of advanced research reactor with high neutron performance has been performed at KAERI based on design and operation experience obtained from HANARO. In this study, nuclear characteristics of design parameters such as various types of fuel assemblies, structural materials of core and fuel assembly, and the number of absorber rods were analyzed. Among rod, plate and tube type fuel assemblies considered, tube type assembly seems to be preferable as a high performance research reactor fuel because of high thermal margin and neutron flux in reflector. Aluminium block as a structural material of core was shown to be superior to flow tube due to higher reactivity and thermal flux in reflector. The stiffener to fix plates in th fuel assembly had the no impact on fast flux in central trap. The reduction of thermal flux in reflector caused by the stiffener was about 7%. If the control absorber rods of 4 mm thickness were chosen, it would be possible to operate the reactor with fresh fuel assemblies from the initial core

  5. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

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

  6. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  7. Particle bed reactor propulsion vehicle performance and characteristics as an orbital transfer rocket

    International Nuclear Information System (INIS)

    Horn, F.L.; Powell, J.R.; Lazareth, O.W.

    1986-01-01

    The particle bed reactor designed for 100 to 300 MW power output using hydrogen as a coolant is capable of specific impulses up to 1000 seconds as a nuclear rocket. A single space shuttle compatible vehicle can perform extensive missions from LEO to 3 times GEO and return with multi-ton payloads. The use of hydrogen to directly cool particulate reactor fuel results in a compact, lightweight rocket vehicle, whose duration of usefulness is dependent only upon hydrogen resupply availability. The LEO to GEO mission had a payload capability of 15.4 metric tons with 3.4 meters of shuttle bay. To increase the volume limitation of the shuttle bay, the use of ammonia in the initial boost phase from LEO is used to give greater payload volume with a small decrease in payload mass, 8.7 meters and 12.7 m-tons. 5 refs., 15 figs

  8. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  9. Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

    1992-01-01

    The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

  10. Principles, design and fuel performance characteristics of gas cooled thermal reactors

    International Nuclear Information System (INIS)

    Boocock, P.M.; Eaton, J.R.P.

    1989-01-01

    Reactor output and availability are closely related to fuel design and performance and the SSEB, in collaboration with the Central Electricity Generating Board have followed a policy of continuous analysis and improvement. The position reached is set out and some views on further improvements, are given. The strategy of increasing fuel burn-up on Hunterston A power station has brought significant dividends in the form of major benefits in fuel cycle cost and station availability. Significant improvements in output and availability at Hunterston B have resulted from increasing the fuel cycle burn-up, from 18 GWd/t U to 21 GWd/t U and introducing on-load refuelling. Additional benefits are soon to be obtained by further extending the burn-up to 24 GWd/t U. Further reduction of typically Pound 2-7 million/year in fuel cycle costs over the remaining life of the stations would be made by extending the burn-up to 30 GWd/t U at Hunterston B and Torness. There would be additional savings of about Pound 4 million/year in replacement fuel costs if the reactors continued to be refuelled at 30% power at Hunterston B and 40% power at Torness. (author)

  11. Deformation Characteristics and Sealing Performance of Metallic O-rings for a Reactor Pressure Vessel

    Directory of Open Access Journals (Sweden)

    Mingxue Shen

    2016-04-01

    Full Text Available This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV. A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

  12. Deformation characteristics and sealing performance of metallic-O-ring for a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ming Xue; Peng, Xudong; Xie, Linjun; Meng, Xiang Kai [Engineering Research Center of Process Equipment and Its Remanufacture, Ministry of Education, Zhejiang University of Technology, Hangzhou (China); Li, Xing Gen [Ningbo Tiansheng Sealing Packing Co., Ltd., Ningbo (China)

    2016-04-15

    This paper provides a reference to determine the seal performance of metallic O-rings for a reactor pressure vessel (RPV). A nonlinear elastic-plastic model of an O-ring was constructed by the finite element method to analyze its intrinsic properties. It is also validated by experiments on scaled samples. The effects of the compression ratio, the geometrical parameters of the O-ring, and the structure parameters of the groove on the flange are discussed in detail. The results showed that the numerical analysis of the O-ring agrees well with the experimental data, the compression ratio has an important role in the distribution and magnitude of contact stress, and a suitable gap between the sidewall and groove can improve the sealing capability of the O-ring. After the optimization of the sealing structure, some key parameters of the O-ring (i.e., compression ratio, cross-section diameter, wall thickness, sidewall gap) have been recommended for application in megakilowatt class nuclear power plants. Furthermore, air tightness and thermal cycling tests were performed to verify the rationality of the finite element method and to reliably evaluate the sealing performance of a RPV.

  13. Material and geometry options and performance characteristics for a test reactor

    International Nuclear Information System (INIS)

    Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

    1993-01-01

    For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

  14. dynamic performance of research reactors

    International Nuclear Information System (INIS)

    Abo elnor, A.G.M.

    2007-01-01

    this work studies the dynamic performance of material testing reactor (MTR), where the dynamic performance of any reactor reflects its safety behavior and it should enhance its intrinsic characteristics s ystem corrects itself internally without introducing external corrective action . the present work analyzes and studies the dynamic performance of mtr through the transfer function. the servo system parameters can be changed to fit the system demand. the servo system is an excellent approximation to some of the practical servo system currently use in reactor control system, and a quadratic form of this sort should closely approximate the behavior of almost any type of physical equipment which might be chosen to drive a control rod. proposed changes in servo system parameters could enhance the dynamic performance of the system , but the suitable parameters can be evaluated by using the automatic reactor power control system model

  15. Start-up and performance characteristics of a trickle bed reactor degrading toluene

    Directory of Open Access Journals (Sweden)

    Ondrej Misiaczek

    2007-09-01

    Full Text Available The objective of this work was to evaluate toluene degradation in a trickle bed reactor when the loading was carried out by changing the air flow rate. The biofiltration system was inoculated with a mixed microbial population, adapted to degradation of hydrophobic compounds. Polypropylene high flow rings were used as a packing material. The system was operated for a period of 50 days at empty bed residence times ranging from 106s to 13s and with a constant inlet concentration of toluene of 100 mg.m-3. The reactor showed high removal efficiency at higher contact times and increasing elimination capacity with higher air-flow rates. The highest EC value reached was 9.8 gC.m-3.h-1 at EBRT = 13s. During the experiment, the consumption of NaOH solution was also measured. No significant variation of this value was found and an average value of 3.84 mmol of NaOH per gram of consumed carbon was recorded.

  16. Reactor core performance calculating device

    International Nuclear Information System (INIS)

    Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.

    1995-01-01

    The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)

  17. Performance characteristics of specified power reactors in multidimensional neutron diffusion problems

    International Nuclear Information System (INIS)

    Kim, M.G.

    1980-01-01

    The multigroup neutron diffusion equations with the constraint of specified power distributions are investigated by the application of the straight-line method which can be considered as the limiting case of zero mesh space in the finite difference method. The standard partial differential form of the diffusion equation is reduced to sets of ordinary differential equations and then converted into sets of integral equations by using Green's functions defined on the pseudo straight lines. Coupling of each straight line to the adjacent lines arises from the application of a three-point central difference formula. The interfaces encountered between two regions are taken into account by imposing the continuity conditions for the grown fluxes and net currents with Taylor expansions of internal fluxes at the interface positions. A few sample problems are selected to test the validity of the method. It is found that the proposed method of solution is similar to the finite Fourier sine transform. Numerical results for the solutions obtained by the method of straight lines are compared with the results of the exact analytical solutions for simple geometries. These comparisons indicate that the proposed method is compatible with the analytical method, and in some problems considered the straight-line solutions are much more efficient than the exact solutions. The method is also extended to the reactor kinetics problem by expressing the kinetics parameters in terms of the basis functions which are used to obtain the solutions of the steady-state neutron diffusion equations

  18. HTR characteristics affecting reactor physics

    International Nuclear Information System (INIS)

    Ehlers, K.

    1980-01-01

    A physical description of high-temperature has-cooled reactors is given, followed by an overview of HTR characteristics. The emphasis is placed on the HTR fuel cycle alternatives and thermohydraulics of pebble bed core. Some prospects of HTRs in the Federal Republic of Germany are also presented

  19. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

    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.

  20. A task-based parallelism and vectorized approach to 3D Method of Characteristics (MOC) reactor simulation for high performance computing architectures

    Science.gov (United States)

    Tramm, John R.; Gunow, Geoffrey; He, Tim; Smith, Kord S.; Forget, Benoit; Siegel, Andrew R.

    2016-05-01

    In this study we present and analyze a formulation of the 3D Method of Characteristics (MOC) technique applied to the simulation of full core nuclear reactors. Key features of the algorithm include a task-based parallelism model that allows independent MOC tracks to be assigned to threads dynamically, ensuring load balancing, and a wide vectorizable inner loop that takes advantage of modern SIMD computer architectures. The algorithm is implemented in a set of highly optimized proxy applications in order to investigate its performance characteristics on CPU, GPU, and Intel Xeon Phi architectures. Speed, power, and hardware cost efficiencies are compared. Additionally, performance bottlenecks are identified for each architecture in order to determine the prospects for continued scalability of the algorithm on next generation HPC architectures.

  1. Identification of nuclear reactor characteristics by the reactor noise analysis

    International Nuclear Information System (INIS)

    Yashima, Hideyuki

    1980-01-01

    Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

  2. Characteristics and performance of aerobic algae-bacteria granular consortia in a photo-sequencing batch reactor.

    Science.gov (United States)

    Liu, Lin; Zeng, Zhichao; Bee, Mingyang; Gibson, Valerie; Wei, Lili; Huang, Xu; Liu, Chaoxiang

    2018-05-05

    The characteristics and performance of algae-bacteria granular consortia which cultivated with aerobic granules and targeted algae (Chlorella and Scenedesmus), and the essential difference between granular consortia and aerobic granules were investigated in this experiment. The result indicated that algae-bacteria granular consortia could be successfully developed, and the algae present in the granular consortia were mainly Chlorella and Scenedesmus. Although the change of chlorophyll composition revealed the occurrence of light limitation for algal growth, the granular consortia could maintain stable granular structure, and even showed better settling property than aerobic granules. Total nitrogen and phosphate in the algal-bacterial granular system showed better removal efficiencies (50.2% and 35.7%) than those in the aerobic granular system (32.8% and 25.6%) within one cycle (6 h). The biodiesel yield of aerobic granules could be significantly improved by algal coupled process, yet methyl linolenate and methyl palmitoleate were the dominant composition of biodiesel obtained from granular consortia and aerobic granules, respectively. Meanwhile, the difference of dominant bacterial communities in the both granules was found at the order level and family level, and alpha diversity indexes revealed the granular consortia had a higher microbial diversity. Copyright © 2018. Published by Elsevier B.V.

  3. Inherent safety characteristics of innovative reactors

    International Nuclear Information System (INIS)

    Heil, J.A.

    1995-11-01

    The added safety value of innovative or third generation reactor designs has been evaluated in order to determine the most suitable candidate for Dutch government funded research and development support. To this end, four innovative reactor concepts, viz. PIUS (Process Inherent Ultimate Safety), PRISM (Power Reactor Innovative Small), HTR-M (High Temperature Reactor Module) and MHTGR (Modular High Temperature Gas-cooled Reactor), have been studied and their passive and inherent safety characteristics have been outlined. Also the outlook for further technological and industrial development has been considered. The results of the study confirm the perspective of the innovative reactors for reduced dependence on active safety provisions and for a further reduced vulnerability to technical failures and human errors. The accident responses to generic accident initiators, viz. reactivity and cooling accidents, and also to reactor specific accidents show that neither active safety systems nor short term operator actions are required for maintaining the reactor core in a controlled and coolable condition. Whether this gives rise to a higher total safety of the innovative reactor designs, compared to evolutionary or advanced reactors, cannot be concluded. Supplementary experimental and analytical analyses of reactor specific accidents are required to be able to assess the safety of these innovative designs in a more quantitative manner. It is believed that the safety case of innovative reactors, which are less dependent on active safety systems, can be communicated with the general public in a more transparent way. Considering the perspective for further technological and industrial development it is not expected that any of the considered innovative reactor concepts will become commercially available within the next one to two decades. However, they could be made available earlier if they would receive sufficient financial backing. Considering the added safety perspectives

  4. Reactor core performance estimating device

    International Nuclear Information System (INIS)

    Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

    1995-01-01

    The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

  5. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

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

    1985-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  7. A Study on the Kinetic Characteristics of Transmutation Process Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae seon; Huh, Chang Wook; Kim, Doh Hyung [Seoul National University, Seoul (Korea, Republic of)

    1997-07-01

    The purpose of this study is to examine the transient heat transfer characteristics of liquid mental as the coolant used in accelerator-driven transmutation process reactor which is related the disposal of high-level radioactive nuclide. At current stage, the accelerator-driven transmutation process is investigated as the most appropriate method among many transmutation process methods. In this study, previous research works are investigated especially about the thermal hydraulics and kinetic behavior of coolant material including heat transfer of coolant in transmutation process reactor. A study on the heat transfer characteristics of liquid metal is performed based on the thermal hydraulic kinetic characteristics of liquid metal reactor which uses liquid metal coolant. Based on this study, the most appropriate material for the coolant of transmutation reactor will be recommended. 53 refs., 15 tabs., 33 figs. (author)

  8. Performance and main characteristic parameters of the Cairo fourier diffractometer facility at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Maayouf, R.M.A.; Abdel-Latif, I.; El-Kady, A.; El-Shafey, A.; Khalil, M.; El-Shaer, Y.

    1997-05-01

    This report represents the results of measurements performed recently with the Cairo Fourier diffractometer facility (CFDF). The main components of the CFDF were supplied by the IAEA according to the technical assistance project EGY/1/022. The CFDF performance is assessed and the main parameters are given. The neutron guide system attached to the CFDF provides a thermal neutron flux ∼ 10 6 n/cm 2 .sec at the sample position; free from fast neutrons and gamma rays background. It has been found, from measurements with different powder samples, that such value of the thermal neutron flux is adequate for neutron diffraction measurements, at scattering angle 2θ 90 deg. and D values between 0.7A and 2.5A; with 52% resolution. (author). 26 refs, 10 figs, 2 tabs

  9. Zero energy reactor RB technical characteristics and experimental possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

  10. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

  11. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  12. Numerical investigation of ethanol fuelled HCCI engine using stochastic reactor model. Part 2: Parametric study of performance and emissions characteristics using new reduced ethanol oxidation mechanism

    International Nuclear Information System (INIS)

    Maurya, Rakesh Kumar; Akhil, Nekkanti

    2016-01-01

    Highlights: • Newly developed reduced ethanol mechanism (47 species and 272 reactions) used. • Engine maps over wide range are developed for performance and emissions parameters. • HCCI operating range increases with compression ratio & decreases with engine speed. • Maximum combustion efficiency up to 99% and thermal efficiency up to 50% is achieved. • Maximum N_2O emission found up to 2.7 ppm and lower load have higher N_2O emission. - Abstract: Ethanol fuelled homogenous charge compression ignition engine offers a better alternative to tackle the problems of achieving higher engine efficiency and lower emissions using renewable fuel. Present study computationally investigates the HCCI operating range of ethanol at different compression ratios by varying inlet air temperature and engine speed using stochastic reactor model. A newly developed reduced ethanol oxidation mechanism with NO_x having 47 species and 272 reactions is used for simulation. HCCI operating range for compression ratios 17, 19 and 21 are investigated and found to be increasing with compression ratio. Simulations are conducted for engine speeds ranging from 1000 to 3000 rpm at different intake temperatures (range 365–465 K). Parametric study of combustion and emission characteristics is conducted and engine maps are developed at most efficient inlet temperatures. HCCI operating range is defined using combustion efficiency (>85%) and maximum pressure rise rate (<5 MPa/ms). In HCCI operating range, higher efficiency is found at higher engine loads and lower engine speeds. Emission characteristics of species (NO_x, N_2O, CO, CH_4, C_2H_4, C_2H_6, CH_3CHO, and HCHO) found in significant amount is also analysed for ethanol fulled HCCI engine. Emission maps for different species are presented and discussed for wide range of speed and load conditions. Some of unregulated species such as aldehydes are emitted in significantly higher quantities from ethanol fuelled HCCI engine at higher load

  13. Inherently safe characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This report is based on a detailed study which was carried out by Colenco (a company of the Motor-Columbus Group) on behalf of the Commission of the European Communities (CEC). It presents a summary of this study and concentrates more on the generic issues involved in the subject of inherent safety in nuclear power plants. It is assumed that the reader is reasonably familiar with the design outline of the systems included in the report. The report examines the role of inherent design features in achieving the safety of nuclear power plants as an alternative to the practice, which is largely followed in current reactors, of achieving safety by the addition of engineered safety features. The report examines current reactor systems to identify the extent to which their characteristics are either already inherently safe or, on the other hand, have inherent characteristics that require protective action to be taken. It then considers the advantages of introducing design changes to improve their inherent safety characteristics. Next, it looks at some new reactor types for which claims of inherent safety are made to see to what extent these claims are justified. The general question is then considered whether adoption of the inherently safe reactors would give advantages (by reducing risk in real terms or by improving the public acceptability of nuclear power) which are sufficient to offset the expected high costs and the technical risks associated with any new technology

  14. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  15. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    Gillies, C.; White, M.

    1995-11-01

    A review of Canadian and worldwide performance indicator definitions and data was performed to identify a set of indicators that could be used for comparison of performance among nuclear power plants. The results of this review are to be used as input to an AECB team developing a consistent set of performance indicators for measuring Canadian power reactor safety performance. To support the identification of performance indicators, a set of criteria was developed to assess the effectiveness of each indicator for meaningful comparison of performance information. The project identified a recommended set of performance indicators that could be used by AECB staff to compare the performance of Canadian nuclear power plants among themselves, and with international performance. The basis for selection of the recommended set and exclusion of others is provided. This report provides definitions and calculation methods for each recommended performance indicator. In addition, a spreadsheet has been developed for comparison and trending for the recommended set of indicators. Example trend graphs are included to demonstrate the use of the spreadsheet. (author). 50 refs., 11 tabs., 3 figs

  16. Characteristics of outage radiation fields around various reactor components

    International Nuclear Information System (INIS)

    Verzilov, Y.; Husain, A.; Corbin, G.

    2008-01-01

    Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

  17. MAPLE research reactor beam-tube performance

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Gillespie, G.E.

    1989-05-01

    Atomic Energy of Canada Limited (AECL) has been developing the MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor concept as a medium-flux neutron source to meet contemporary research reactor applications. This paper gives a brief description of the MAPLE reactor and presents some results of computer simulations used to analyze the neutronic performance. The computer simulations were performed to identify how the MAPLE reactor may be adapted to beam-tube applications such as neutron radiography

  18. High performance light water reactor

    International Nuclear Information System (INIS)

    Squarer, D.; Schulenberg, T.; Struwe, D.; Oka, Y.; Bittermann, D.; Aksan, N.; Maraczy, C.; Kyrki-Rajamaeki, R.; Souyri, A.; Dumaz, P.

    2003-01-01

    The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

  19. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  20. Safety characteristics of small heat producing reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1987-10-01

    The primary objectives of protection in nuclear power plants are the possibility to shut the reactor down in case of emergency and keep it subcritical in the long run, the existence of a heat sink for post-decay heat removal in order to avoid overheating, let alone core meltdown, and the containment of radioactivity within the barriers designed for this purpose, thus preventing significant activity release. In principle, these objectives can be met in various ways, namely by active, passive or inherent technical safeguards systems. In practice, a mixture of these approaches is employed in almost all cases. What matters in the end is the assessment of the overall concept, not of some outstanding feature. Inherent characteristics are easier to achieve in small reactors. However, also in this case, inherent safety does not mean absolute safety. If inherent safety characteristics were all encompassing, they would have to include self-healing effects. However, inanimate matter is incapable of such self-organization. Consequently, inherent characteristics in nuclear technology by definition should include the increased use of dissipative processes in the thermal part of the plant. (author)

  1. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  2. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  3. Transient thermal characteristics of a core channel in a molten salt reactor

    International Nuclear Information System (INIS)

    Sakashita, H.; Ishiguro, R.; Sugiyama, K.

    1987-01-01

    The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

  4. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  5. Neutronics characteristics of space power reactors

    International Nuclear Information System (INIS)

    Little, W.; Barner, J.

    1986-01-01

    The objective of the paper is to describe the neutronic characteristics of a range of possible space reactor designs, and indicate the relative advantages and disadvantages of the various designs. Fuel designs to be considered are cermets (i.e., ceramic particles embedded in a metal matrix) consisting of UO 2 or Nn ceramic particles in matrices of Nb, Mo, Ta, or W. These cermet fuels are compared to a UN pin-type design. UN was selected for the reference fuel material since it has a somewhat higher density than UO 2 (i.e., 14.32 versus 10.96 gm/cc), which allows a lower minimum critical mass for both ceramic and cermet designs

  6. Expected characteristics of future reactors for human beings

    International Nuclear Information System (INIS)

    Taketani, Kiyoaki

    1992-01-01

    Based on four reactor safety components (namely: a) God-given safety, b) Equipment safety, c) Quick-response safety, d) Containing safety), categorical assessment is made of various nuclear reactor concepts ranging from present existing reactors to future reactors based on innovative reactor design. In pursuit of nuclear reactor safety, ultimate characteristics of the ideal nuclear reactor are expected to coincide with those of an inherently safe reactor. A definition of 'inherently safe' has already been proposed by a committee in Japan. As a realistic and existable reactor, which is as close to the ideal reactor, a future reactor which is almost the same as a global reactor, is proposed. This global reactor must be constructable anywhere on earth and must permit easy operation and maintenance by anyone. It is also discussed to identify what behavior is expected of the global reactor under various conditions. At the same time, this future reactor which includes the global reactor, should solve a) the nuclear fuel resource issue, b) efficient utilization of fission energy and c) environmental issues as the greenhouse effect. (author). 7 refs., 2 figs

  7. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  8. Optimization method development of the core characteristics of a fast reactor in order to explore possible high performance solutions (a solution being a consistent set of fuel, core, system and safety)

    International Nuclear Information System (INIS)

    Ingremeau, J.-J.X.

    2011-01-01

    In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc...) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) uses analytical models and interpolations (Meta-models) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a 'buffer', a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that

  9. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

  10. Performance Characteristics of the Experimental Boiling Water Reactor from 0 to 100 MW(t); Performances de l'EBWR de 0 a 100 MW; Rabochaya kharakteristika ehksperimental'nogo kipyashchego reaktora EBWR pri moshchnosti 0 - 100 mgvt.; Rendimiento del reactor experimental de agua hirviente (EBWR) entre 0 y 100 MW

    Energy Technology Data Exchange (ETDEWEB)

    Iskenderian, A.; Lipinski, W. C.; Petrick, M.; Wimunc, E. A. [Argonne National Laboratory, Argonne, IL (United States)

    1963-10-15

    On 25 May 1962 the Argonne National Laboratory received approval from the USAEC to operate EBWR to a power level of 100 MW. Administrative approval to proceed was granted by the International Atomic Energy Agency safeguards system on 11 July 1962. On 15 November 1962 an operating power level of 100 MW was reached. The EBWR 100 MW Reactor Experimental Program was completed on 6 December 1962. One of the major goals of this project was to instrument the reactor extensively in order to obtain data and information on the performance characteristics of this reactor type. The programme was the first of its kind to be undertaken and the first to be completed. Many new instrumentation techniques had to be developed for obtaining the desired data. The goal was successfully achieved, and many new data were obtained on the performance characteristics of a natural circulation boiling-water reactor. The data derived from this programme provided information on recirculation flow rates, vapour liquid separation limits (steam carryunder in the downcomer and liquid carryover in the effluent steam), subcooling, location oi the true interface in the reactor and its relation to the water column level, steam collapse rates in the downcomer, void coefficients, H{sub 3}BO{sub 3} worth, temperature coefficients, use of boron strips for control purposes, use of spike elements, transfer functions, noise analysis, some flux measurements, stability, etc. In addition, data were obtained on the behaviour and integrity of certain reactor components and systems such as boric acid control, radiation levels, corrosion product distribution, equipment malfunctions, fuel and control rods, etc. EBWR's performance characteristics were governed almost exclusively by steam carryunder in the downcomer, liquid carryover in the effluent steam, and indirectly by the location of the true interface in the vessel. Carryunder was the dominating factor in the lower power range. Above 65 MW the reactor

  11. The Performance of Structured Packings in Trickle-Bed Reactors

    NARCIS (Netherlands)

    Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped

  12. Reactor shutdown: nuclear power plant performance

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The article essentially looks at the performance of nine of Sweden's nuclear reactors. A table lists the percentage of time for the first three quarters of 1981 that the reactors were operating, and the number of hours out of service for planned or other reasons. In particular, one station - Ringhals 3 - was out of action because of a damaged tube in the associated steam generator. The same fault occurred with another reactor - Ringhals 4 - before this was brought into service. The reasons for the failure and its importance are briefly discussed. (G.P.)

  13. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  14. Noise analysis of the Dodewaard boiling water reactor: characteristics and time history

    International Nuclear Information System (INIS)

    Veer, J.H.C. v.d.; Kema, N.V.

    1982-01-01

    Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)

  15. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  16. Performance Indicators of Operating Reactors

    Data.gov (United States)

    Nuclear Regulatory Commission — A list of Performance Indicators (PI) that are reported to the NRC by licensees at the end of each quarter in accordance with Inspection Manual Chapters (IMC) 0608,...

  17. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  18. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-05-01

    Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K to 1313 K, under atmospheric pressure. Sole distinctive weak flame was observed for each mixture, with inlet fuel/air mixture velocity set low at 2 cm/s. One-dimensional computation with comprehensive chemistry and transport was conducted. At low mixture velocities, one-stage oxidation was confirmed from heat release rate profiles, which was broadly in agreement with the experimental results. The weak flame positions were congruent with literature describing reactivity of the butanol isomers. These weak flame responses were also found to mirror the trend in Anti-Knock Indexes of the butanol isomers. Flux and sensitivity analyses were performed to investigate the fuel oxidation pathways at low and high temperatures. Further computational investigations on oxidation of butanol isomers at higher pressure of 5 atm indicated two-stage oxidation through the heat release rate profiles. Low temperature chemistry is accentuated in the region near the first weak cool flame for oxidation under higher pressure, and its impact on key species – such as hydroxyl radical, hydrogen peroxide and carbon monoxide – were considered. Both experimental and computational findings demonstrate the advantage of employing the micro flow reactor in investigating oxidation processes in the temperature region of interest along the reactor channel. By varying physical conditions such as pressure, the micro flow reactor system is proven to be highly beneficial in elucidating oxidation behavior of butanol isomers in conditions in engines such as those that mirror HCCI operations.

  19. Study and application of boiling water reactor jet pump characteristic

    International Nuclear Information System (INIS)

    Liao Lihyih

    1992-01-01

    RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

  20. Physical Characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy

    1994-10-01

    The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

  1. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  2. Evaluation of the influence of seismic restraint characteristics on breeder reactor piping systems

    International Nuclear Information System (INIS)

    Mello, R.M.; Pollono, L.P.

    1979-01-01

    For the Clinch River Breeder Reactor Plant (CRBRP) heat transport system piping within the reactor containment building, dynamic analyses of the piping loops have been performed to study the effect of restraint stiffness on the dynamic behavior of the piping. In addition, analysis and testing of typical CRBRP restraint system components have been performed for the purpose of quantifying and verifying the basic characteristics of the restraints used in the piping system dynamic analysis

  3. The effect of hydraulic retention time on the performance and fouling characteristics of membrane sequencing batch reactors used for the treatment of synthetic petroleum refinery wastewater.

    Science.gov (United States)

    Shariati, Seyed Ramin Pajoum; Bonakdarpour, Babak; Zare, Nasim; Ashtiani, Farzin Zokaee

    2011-09-01

    The use of membrane sequencing batch reactors, operated at HRT of 8, 16 and 24h, was considered for the treatment of a synthetic petroleum wastewater. Increase in HRT resulted in statistically significant decrease in MLSS. Removal efficiencies higher than 97% were found for the three model hydrocarbon pollutants at all HRTs, with air stripping making a small contribution to overall removal. Particle size distribution (PSD) and microscopic analysis showed reduction in the protozoan populations in the activated sludge with decreasing HRT. PSD analysis also showed a higher proportion of larger and smaller sized particles at the lowest HRT. The rate of membrane fouling was found to increase with decreasing HRT; SMP, especially carbohydrate SMP, and mixed liquor apparent viscosity also showed a pronounced increase with decreasing HRT, whereas the concentration of EPS and its components decreased. FTIR analysis identified organic compounds as the main component of membrane pore fouling. Copyright © 2011 Elsevier Ltd. All rights reserved.

  4. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  5. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  6. Research reactor RB, technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1978-01-01

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) [sr

  7. Passive safety design characteristics of the KALIMER-600 burner reactor

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

    2009-01-01

    The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

  8. Economic characteristics of a smaller, simpler reactor

    International Nuclear Information System (INIS)

    LaBar, M.; Bowers, H.

    1988-01-01

    Reduced load growth and heightened concern with economic risk has led to an expressed utility preference for smaller capacity additions. The Modular High Temperature Reactor (MHTGR) plant has been developed as a small, simple plant that has limited financial risk and is economically competitive with comparatively sized coal plants. Competitive economics is achieved by the simplifications made possible in a small MHTGR, reduction in the quantity of nuclear grade construction and design standardization and certification. Assessments show the MHTGR plant to have an economic advantage over coal plants for plant sizes from 270 MWe to 1080 MWe. Financial risk is limited by small unit sizes and short lead times that allow incremental deployment. Evaluations show the MHTGR incremental deployment capability to reduce negative cash flows by almost a factor of 2 relative to that required by a single large nuclear plant

  9. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    Heindler, M.; Harms, A.A.

    1978-01-01

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  10. Safety characteristics of the integral fast reactor concept

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

    1985-01-01

    The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

  11. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  12. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-01-01

    Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K

  13. Kinetic characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    An, Tran Khac; Dien, Nguyen Nhi; Hien, Pham Duy [Nuclear Research Inst., Da Lat (Viet Nam); and others

    1994-10-01

    Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be({gamma}, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs.

  14. Kinetic characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Tran Khac An; Nguyen Nhi Dien; Pham Duy Hien

    1994-01-01

    Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be(γ, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs

  15. DUPIC fuel performance from reactor physics viewpoint

    International Nuclear Information System (INIS)

    Choi, H.; Rhee, B.W.; Park, H.

    1995-01-01

    A preliminary study was performed for the evaluation of Stress Corrosion Cracking (SCC) parameters of nominal DUPIC fuel in CANDU reactor. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increase of the 43-element DUPIC fuel in the equilibrium core are below the SCC thresholds of CANDU natural uranium fuel. For 4-bundle shift refueling scheme, the envelope of element ramped power and power increase upon refueling are 8% and 44% higher than those of 2-bundle shift refueling scheme on the average, respectively, and both schemes are not expected to cause SCC failures. (author)

  16. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  17. Reactor fuel performance data file, 1985 edition

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Fujita, Misao; Watanabe, Kohji.

    1986-07-01

    In safety evaluation and integrity studies of reactor fuel, data on fuel performance are the most basic materials. The Fuel Reliability Laboratory No.1 has obtained the fuel performance data by joining in some international programs to study the safety and integrity of fuel. Those data have only used for the studies in the above two fields. However, if the data are rearranged and compiled in a easily usable form, they can be utilized in other field of studies. Then, a 'data file' on fuel performance is beeing compiled by adding data from open literatures to those obtained in international programs. The present report is prepared on the basis of the data file compiled by March in 1986. (author)

  18. Characteristics of a reactor with power reactivity feedback

    International Nuclear Information System (INIS)

    Li Fengyu; Zhang Yusheng; Zhang Guangfu; Liu Ying

    2008-01-01

    The point-reactor model with power reactivity feedback becomes a nonlinear system. Its dynamic characteristic shows great complexity. According to the mathematic definition of stability in differential equation qualitative theory, the model of a reactor with power reactivity feedback is judged unstable. The equilibrium point is a saddle-node point. A portion of the trajectory in the neighborhood of the equilibrium point is parabolic fan curve, and the other is hyperbolic fan curve. Based on phase locus near the equilibrium point, it is pointed out that the model is still stable within physical limits. The difference between stabilities in the mathematical sense and in the physical sense is indicated. (authors)

  19. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  20. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  1. Fabrication of high performance components for Indian nuclear reactors

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2011-01-01

    Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

  2. The need for high performance breeder reactors

    International Nuclear Information System (INIS)

    Vaughan, R.D.; Chermanne, J.

    1977-01-01

    It can be easily demonstrated, on the basis of realistic estimates of continued high oil costs, that an increasing portion of the growth in energy demand must be supplied by nuclear power and that this one might account for 20% of all the energy production by the end of the century. Such assumptions lead very quickly to the conclusion that the discovery, extraction and processing of the uranium will not be able to follow the demand; the bottleneck will essentially be related to the rate at which the ore can be discovered and extracted, and not to the existing quantities nor their grade. Figures as high as 150.000 T/annum and more would be quickly reached, and it is necessary to wonder already now if enough capital can be attracted to meet these requirements. There is only one solution to this problem: improve the conversion ratio of the nuclear system and quickly reach the breeding; this would lead to the reduction of the natural uranium consumption by a factor of about 50. However, this condition is not sufficient; the commercial breeder must have a breeding gain as high as possible because the Pu out-of-pile time and the Pu losses in the cycle could lead to an unacceptable doubling time for the system, if the breeding gain is too low. That is the reason why it is vital to develop high performance breeder reactors. The present paper indicates how the Gas-cooled Breeder Reactor [GBR] can meet the problems mentioned above, on the basis of recent and realistic studies. It briefly describes the present status of GBR development, from the predecessors in the gas cooled reactor line, particularly the AGR. It shows how the GBR fuel takes mostly profit from the LMFBR fuel irradiation experience. It compares the GBR performance on a consistent basis with that of the LMFBR. The GBR capital and fuel cycle costs are compared with those of thermal and fast reactors respectively. The conclusion is, based on a cost-benefit study, that the GBR must be quickly developed in order

  3. The safety characteristics of the HTR 500 reactor plant

    International Nuclear Information System (INIS)

    Wachholz, W.

    1987-01-01

    The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

  4. Primary circuit and reactor core T-H characteristics determination of WWER 440 reactors

    International Nuclear Information System (INIS)

    Hermansky, J.; Petenyi, V.; Zavodsky, M.

    2010-01-01

    The WWER-440 nuclear fuel vendor permanently improves the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. During unit refuelling there also could be made some other changes in hydraulic parameters of primary circuit (change of impeller wheels, hydraulic resistance coefficient changes of internal parts of primary circuit, etc.). Therefore it is necessary to determine real coolant flow rate through the reactor during units start-up after their refuelling, and also to have the skilled methodology and computing code for analyzing factors, which affecting the inaccuracy of coolant flow redistribution determination through reactor on flows through separate parts of reactor core in any case of parallel operation of different assembly types. Computing code TH-VCR and CORFLO are used for reactor core characteristics determination for one type of fuel and control assemblies and also in case of parallel operation of different assembly types. The code TH-VCR is able to calculate coolant flow rate for different combinations of three different fuel assembly channel types and three different control assembly channel types. The CORFLO code deals the area of the reactor core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is taken into account. Computing code CORFLO is verified and validated at this time. Paper presents some results from measurements of coolant flow rate through reactors during start-up after unit refuelling and short description of computing code TH-VCR and CORFLO with some calculated results. (Authors)

  5. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  6. Characteristics of fast reactor core designs and closed fuel cycle

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

    2007-01-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  7. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

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

  8. CHARACTERISTICS OF CORN STALK HEMICELLULOSE PYROLYSIS IN A TUBULAR REACTOR

    OpenAIRE

    Gao-Jin Lv; Shu-Bin Wu; Rui Lou

    2010-01-01

    Pyrolysis characteristics of corn stalk hemicellulose were investigated in a tubular reactor at different temperatures, with focus mainly on the releasing profiles and forming behaviors of pyrolysis products (gas, char, and tar). The products obtained were further identified using various approaches (including GC, SEM, and GC-MS) to understand the influence of temperature on product properties and compositions. It was found that the devolatilization of hemicellulose mainly occurred at low tem...

  9. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  10. Design and performance of subgrade biogeochemical reactors.

    Science.gov (United States)

    Gamlin, Jeff; Downey, Doug; Shearer, Brad; Favara, Paul

    2017-12-15

    Subgrade biogeochemical reactors (SBGRs), also commonly referred to as in situ bioreactors, are a unique technology for treatment of contaminant source areas and groundwater plume hot spots. SBGRs have most commonly been configured for enhanced reductive dechlorination (ERD) applications for chlorinated solvent treatment. However, they have also been designed for other contaminant classes using alternative treatment media. The SBGR technology typically consists of removal of contaminated soil via excavation or large-diameter augers, and backfill of the soil void with gravel and treatment amendments tailored to the target contaminant(s). In most cases SBGRs include installation of infiltration piping and a low-flow pumping system (typically solar-powered) to recirculate contaminated groundwater through the SBGR for treatment. SBGRs have been constructed in multiple configurations, including designs capable of meeting limited access restrictions at heavily industrialized sites, and at sites with restrictions on surface disturbance due to sensitive species or habitat issues. Typical performance results for ERD applications include 85 to 90 percent total molar reduction of chlorinated volatile organic compounds (CVOCs) near the SBGR and rapid clean-up of adjacent dissolved contaminant source areas. Based on a review of the literature and CH2M's field-scale results from over a dozen SBGRs with a least one year of performance data, important site-specific design considerations include: 1) hydraulic residence time should be long enough for sufficient treatment but not too long to create depressed pH and stagnant conditions (e.g., typically between 10 and 60 days), 2) reactor material should balance appropriate organic mulch as optimal bacterial growth media along with other organic additives that provide bioavailable organic carbon, 3) a variety of native bacteria are important to the treatment process, and 4) biologically mediated generation of iron sulfides along with

  11. Developing countries SMEs innovation characteristics and performance

    DEFF Research Database (Denmark)

    Vang, Jan; Rezaei, Shahamak; Baklanov, Nikita

    An econometric study analysing developing countries’ SMEs innovation characteristics and their correlation with performance.......An econometric study analysing developing countries’ SMEs innovation characteristics and their correlation with performance....

  12. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  13. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1986-05-01

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 10 15 n/cm 2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author) [pt

  14. Nuclear characteristics of D-D fusion reactor blankets, (1)

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

    1977-01-01

    Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

  15. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    Kleiss, J.

    1983-01-01

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  16. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  17. Response characteristics of reactor building on weathered soft rock ground

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Tochigi, Hitoshi

    1991-01-01

    The purpose of this study is to investigate the seismic stability of nuclear power plants on layered soft bedrock grounds, focusing on the seismic response of reactor buildings. In this case, the soft bedrock grounds refer to the weathered soft bedrocks with several tens meter thickness overlaying hard bedrocks. Under this condition, there are two subjects regarding the estimation of the seismic response of reactor buildings. One is the estimation of the seismic response of surface ground, and another is the estimation of soil-structure interaction characteristics for the structures embedded in the layered grounds with low impedandce ratio between the surface ground and the bedrock. Paying attention to these subjects, many cases of seismic response analysis were carried out, and the following facts were clarified. In the soft rock grounds overlaying hard bedrocks, it was proved that the response acceleration was larger than the case of uniform hard bedrocks. A simplified sway and rocking model was proposed to consider soil-structure interaction. It was proved that the response of reactor buildings was small when the effect of embedment was considered. (K.I.)

  18. A model to describe the performance of the UASB reactor.

    Science.gov (United States)

    Rodríguez-Gómez, Raúl; Renman, Gunno; Moreno, Luis; Liu, Longcheng

    2014-04-01

    A dynamic model to describe the performance of the Upflow Anaerobic Sludge Blanket (UASB) reactor was developed. It includes dispersion, advection, and reaction terms, as well as the resistances through which the substrate passes before its biotransformation. The UASB reactor is viewed as several continuous stirred tank reactors connected in series. The good agreement between experimental and simulated results shows that the model is able to predict the performance of the UASB reactor (i.e. substrate concentration, biomass concentration, granule size, and height of the sludge bed).

  19. The study on the mechanical characteristics of concrete of nuclear reactor containment structure

    International Nuclear Information System (INIS)

    Jung, W. S.; Kwon, K. J.; Cho, M. S.; Song, Y. C.

    2000-01-01

    Reactor containment structure of nuclear power plant designed by prestressed concrete causes time-dependent prestress loss due to the mechanical characteristics of concrete. Prestress loss strongly affects to the safety factor of structure under the circumstances of designing, construction and inspection. Thus, this study is to investigate the mechanical characteristics of reactor containment concrete structure of Yonggwang No. 5 and 6. In this study, the compressive strength, modulus of elasticity, poisson's ratio and creep test followed by ASTM code are performed to investigate the mechanical characteristics of concrete made by V type cement. Additionally, since creep causes more time-dependent prestress loss than the other, the measurement value from the creep test is compared with the results from the creep prediction equations by KSCE, JSCE, Hansen, ACI and CEB-FIP model for the effective application. Hereafter, the results of this study may enable to assist the calculation effective stress considering time-dependent prestress loss of the prestressed concrete structures

  20. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  1. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  2. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  3. Performance of a multipurpose research electrochemical reactor

    International Nuclear Information System (INIS)

    Henquin, E.R.; Bisang, J.M.

    2011-01-01

    Highlights: → For this reactor configuration the current distribution is uniform. → For this reactor configuration with bipolar connection the leakage current is small. → The mass-transfer conditions are closely uniform along the electrode. → The fluidodynamic behaviour can be represented by the dispersion model. → This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of ±10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  4. Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

    International Nuclear Information System (INIS)

    Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

    2014-01-01

    A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

  5. Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

    International Nuclear Information System (INIS)

    Heidet, Florent; Kim, Taek K.; Taiwo, Temitope A.

    2011-01-01

    Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

  6. Flow characteristics of Korea multi-purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heonil Kim; Hee Taek Chae; Byung Jin Jun; Ji Bok Lee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-09-01

    The construction of Korea Multi-purpose Research Reactor (KMRR), a 30 MW{sub th} open-tank-in-pool type, is completed. Various thermal-hydraulic experiments have been conducted to verify the design characteristics of the KMRR. This paper describes the commissioning experiments to determine the flow distribution of KMRR core and the flow characteristics inside the chimney which stands on top of the core. The core flow is distributed to within {+-}6% of the average values, which is sufficiently flat in the sense that the design velocity in the fueled region is satisfied. The role of core bypass flow to confine the activated core coolant in the chimney structure is confirmed.

  7. Oscillation characteristics of the reactor 'A'; Oscilatorne karakteristike reaktora 'A'

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Lolic, B [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1961-07-01

    In addition to good knowledge of reactor physical properties, design of the reactor oscillator demands determining of the oscillator operating points as well as oscillation reactor properties. This paper contains study of the RA reactor power changes due to oscillations in in one of the vertical experimental channels. It has been concluded that the reactor optimum operating conditions are attained when the oscillator operates at optimum points, and other parameters are determined dependent on the sensitivity of the method and reactor stability.

  8. Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

    International Nuclear Information System (INIS)

    Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

    1993-01-01

    The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

  9. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  10. Problems in the assessment of inherent safety characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Garribba, S.F.; Vivante, C.

    1988-01-01

    A number of proposals are being made for an increased RD and D effort on advanced nuclear power reactors that would display outstanding safety performance. A common characteristic of the different reactor concepts would be their limited reliance upon active engineered systems under major accident conditions. However, when submitted to a more close scrutiny reactor concept options may reveal diverging safety behaviors and also development opportunities. In this respect, three issues are explored in this paper. A first question is the meaning of non-active, i.e. inherent and passive safety features. Next, is the ranking of advanced and new reactor concepts from the viewpoint of inherent and passive safety. Multiple correspondence analysis may provide a simple tool, whose use is shown for the case of HTR-500, AP600 and PRISM. Conversely, probabilistic risk assessment would allow quantitative comparisons, although lack of information and data is an obstacle. Finally, is demonstration of safety performances as a step toward market deployment of the new reactor systems

  11. Fast pyrolysis of Miscanthus sinensis in fluidized bed reactors: Characteristics of product yields and biocrude oil quality

    International Nuclear Information System (INIS)

    Bok, Jin Pil; Choi, Hang Seok; Choi, Joon Weon; Choi, Yeon Seok

    2013-01-01

    In the present work, fast pyrolysis of Miscanthus sinensis was performed and the product yields and properties of the resulting biocrude oil were determined for varying reactor configurations and pyrolysis temperatures. Two types of reactors (rectangular and cylindrical fluidized beds) were adopted, and pyrolysis temperature was increased from 400 °C to 550 °C. Based on the results, it was found that the reaction temperature greatly influenced the product yield and the characteristics of biocrude oil. The highest yield of biocrude oil for the rectangular reactor was 48.9 wt.%, produced at 500 °C, and the highest yield for the cylindrical reactor was 50.01 wt.%, produced at 450 °C. Additionally, the biocrude oil yield in the rectangular reactor sharply decreased when reaction temperature was increased to 550 °C, while only a slight decrease was observed in the cylindrical reactor. From GC/MS analysis, biocrude oil was found to contain various chemical components, such as nonaromatic ketones, furans, sugars, lignin-derived phenols, guaiacols and syringols. In particular, the sugar content of the biocrude oil produced in rectangular reactor (2.11–9.35 wt.%) was generally lower than that produced in the cylindrical reactor (7.93–10.79 wt.%). - Highlights: • Fast pyrolysis of Miscanthus sinensis was performed in two fluidized bed reactors to obtain biocrude oil. • The yield and characteristics of the biocrude oil were scrutinized with changing reaction temperature and reactor type. • The reaction temperature was found to be the most influencing parameter for the fast pyrolysis reaction. • The different heating rate caused by reactor type has an effect on the final product yield and characteristics

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

    International Nuclear Information System (INIS)

    Buhay, S.

    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.)

  13. Mixing and scale affect moving bed biofilm reactor (MBBR) performance

    NARCIS (Netherlands)

    Kamstra, Andries; Blom, Ewout; Terjesen, Bendik Fyhn

    2017-01-01

    Moving Bed Biofilm Reactors (MBBR) are used increasingly in closed systems for farming of fish. Scaling, i.e. design of units of increasing size, is an important issue in general bio-reactor design since mixing behaviour will differ between small and large scale. Research is mostly performed on

  14. INDIAN POINT REACTOR STARTUP AND PERFORMANCE

    Energy Technology Data Exchange (ETDEWEB)

    Deddens, J. C.; Batch, M. L.

    1963-09-15

    The testing program for the Indian Point Reactor is discussed. The thermal and hydraulic evaluation of the primary coolant system is discussed. Analyses of fuel loading and initial criticality, measurement of operating coefficients of reactivity, control rod group reactivity worths, and xenon evaluation are presented. (R.E.U.)

  15. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

  16. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    Notley, M.J.F.

    1983-07-01

    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 UO 2 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

  17. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  18. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  19. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  20. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  1. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  2. A new high performance research reactor

    International Nuclear Information System (INIS)

    Abbate, Pablo M.

    2002-01-01

    A contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000 between Australia authorities and INVAP from Argentina. Since then the detailed design has been completed, an application for a construction license was made in May 2001 and the construction authorisation was issued on 4 th April 2002. This paper explains the safety philosophy embedded into the design together with the approach taken for main elements of the design and their relation to the proposed applications of the reactor. Also information is provided on the suit of neutron beam facilities and irradiation facilities being constructed. Finally it is presented an outline of the project management organisation, project planing and schedule. (author)

  3. Australia's new high performance research reactor

    International Nuclear Information System (INIS)

    Miller, R.; Abbate, P.M.

    2003-01-01

    A contract for the design and construction of the Replacement Research Reactor was signed in July 2000 between ANSTO and INVAP from Argentina. Since then the detailed design has been completed, a construction authorization has been obtained, and construction has commenced. The reactor design embodies modern safety thinking together with innovative solutions to ensure a highly safe and reliable plant. Also significant effort has been placed on providing the facility with diverse and ample facilities to maximize its use for irradiating material for radioisotope production as well as providing high neutron fluxes for neutron beam research. The project management organization and planing is commensurate with the complexity of the project and the number of players involved. (author)

  4. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

    2000-01-01

    The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

  5. Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-07-01

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

  6. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  7. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  8. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  9. Study on vertical seismic response characteristics of deeply embedded reactor building

    International Nuclear Information System (INIS)

    Morishita, H.; Nakamura, N.; Uchiyama, S.; Fukuoka, A.; Ishizaki, M.

    1993-01-01

    This paper describes vertical response characteristics, especially effects of embedment, and analytical methods for seismic design of a deeply embedded reactor building. The influence of embedment on vertical response was found to be minimal by evaluating results of forced vibration tests of a reactor building model and performing simplified analyses. Subsequently, simulation analyses of the forced vibration test and actual earthquake induced response were performed using both the axisymmetric FEM model and the simplified mass and spring model. It was concluded that the analytical models taking the embedment into the consideration closely simulated the observation records, and the omission of embedment in the analyses tended to increase the predicted response which was conservative in respect an actual design consideration. (author)

  10. Comparison and analysis on transient characteristics of integral pressurized water reactors

    International Nuclear Information System (INIS)

    Zhang, Guoxu; Xie, Heng

    2017-01-01

    Highlights: • Two IPWR Relap5 models with different PSS design were developed. • Postulated SBO and SBLOCA were analyzed. • PRHRS in primary PSS design showed stable performance under different scenarios. • Secondary PRHRS design faced flow instability. - Abstract: In the present work, the similarities and differences of representative IPWRs (integral pressurized water reactor) are studied, and two typical reactor design schemes are summarized. To get a comprehensive understanding of their transient characteristics, SBO (station blackout) and SBLOCA (small break LOCA) are simulated and analyzed respectively by using Relap5/Mod3.2. The calculation results show that, both designs are effective in keeping reactor safe. However, the transient features of the two designs show significant differences. In the primary side passive safety system (PSS) connection design, PRHRS (passive residual heat removal system) shows a roughly congruent performance in removing residual heat under various accidents. While in secondary side PSS connection design, the capability of PRHRS is closely related to primary coolant circulation condition. In SBLOCA analysis, different design approach shows different primary coolant water inventory change trend. And primary PSS connection design could potentially keep reactor core well covered for a longer time.

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  12. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

    1987-01-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  13. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

    Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

  14. Physics methods for calculating light water reactor increased performances

    International Nuclear Information System (INIS)

    Vandenberg, C.; Charlier, A.

    1988-01-01

    The intensive use of light water reactors (LWRs) has induced modification of their characteristics and performances in order to improve fissile material utilization and to increase their availability and flexibility under operation. From the conceptual point of view, adequate methods must be used to calculate core characteristics, taking into account present design requirements, e.g., use of burnable poison, plutonium recycling, etc. From the operational point of view, nuclear plants that have been producing a large percentage of electricity in some countries must adapt their planning to the need of the electrical network and operate on a load-follow basis. Consequently, plant behavior must be predicted and accurately followed in order to improve the plant's capability within safety limits. The Belgonucleaire code system has been developed and extensively validated. It is an accurate, flexible, easily usable, fast-running tool for solving the problems related to LWR technology development. The methods and validation of the two computer codes LWR-WIMS and MICROLUX, which are the main components of the physics calculation system, are explained

  15. Effect of beta limits on reactor performance in EBT

    International Nuclear Information System (INIS)

    Uckan, N.A.; Spong, D.A.; Nelson, D.B.

    1981-01-01

    Because of uncertainties in extrapolating results of simplified models to a reactor plasma, the parameters that influence the beta limits cannot be determined accurately at the present time. Also, the reasonable changes within the models and/or assumptions are seen to affect the core beta limits by almost an order of magnitde. Hence, at the present, these limits cannot be used as a rigid (and reliable) requirement for ELMO Bumpy Torus (EBT) reactor engineering considerations. However, sensitivity studies can be carried out to determine the boundaries of the operating regime and to demonstrate the effects of various modes, assumptions, and models on reactor performance (Q value). First, the modes believed to limit the core β and ring plasma performance are discussed, and the simplifications and/or assumptions involved in deriving these limits are highlighted. Then, the implications of these limits for a reactor are given

  16. Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

    Science.gov (United States)

    Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

    2018-04-01

    1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

  17. Board characteristics, governance objectives, and hospital performance

    DEFF Research Database (Denmark)

    Thiel, Andrea; Winter, Vera; Büchner, Vera Antonia

    2018-01-01

    membership relates to board characteristics and financial performance. METHODOLOGY: Using factor analysis, we identify latent classes of governance objectives and use hierarchical cluster analysis to detect distinct clusters with varying emphasis on the classes. We then use multinomial regression to explore...... the associations between cluster membership and board characteristics (size, gender diversity, and occupational diversity) and examine the associations between clusters and financial performance using OLS regression. RESULTS: Classes of objectives reflecting three governance theories-agency theory, stewardship...... and hospital financial performance, with two of three groups performing significantly better than the reference group. CONCLUSION: High performance in hospitals can be the result of governance logics, which, compared to simple board characteristics, are associated with better financial outcomes. PRACTICE...

  18. Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1997-07-01

    The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

  19. Breeder design for enhanced performance and safety characteristics

    International Nuclear Information System (INIS)

    Fischer, G.J.; Atefi, B.; Yang, J.W.; Galperin, A.; Segev, M.

    1980-01-01

    A fast breeder reactor design has been created which offers a considerably extended fuel cycle and excellent performance characteristics. An example of a core designed to operate on a ten-year fuel cycle is described in some detail. Use of metal fuel along with a moderator such as beryllium oxide dispersed throughout the core provides both design flexibility and safety advantages such as a strong Doppler feedback and limited sodium void reactivity gain. Local power variations are small for the entire cycle; control requirements are also modest, and fuel cycle costs are low

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

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Sunagawa, Takeyoshi

    2014-01-01

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

  1. Effect of Reflector Material on the Neutronic Characteristics of the Small Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sung Hwan; Baek, Min Ho; Yoo, Jae Woon; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The sodium-cooled fast reactor (SFR) has been chosen as a candidate for the Gen-IV Nuclear Energy Systems Initiative due to the advantages in utilization of uranium resources and reduction of radioactive wastes. Recently, the uranium blanket concept is omitted for a purpose of the non-proliferation, hence the reflector material plays a more important role in reactor core design. Moreover, especially in the Korean prototype SFR, the initial core should startup with low-enriched uranium ({<=} 20 w/o) for 100 {approx} 150 MWe power. This restriction causes significant difficulties to achieve sufficient excess reactivity. Thus, in this paper, core characteristic studies of various reflector materials (HT9, BeO, MgO, and ZrH{sub 1.6}) are performed to enhance the initial core excess reactivity

  2. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

  3. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

    1990-01-01

    LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

  4. Growth performance, carcass and hematological characteristics of ...

    African Journals Online (AJOL)

    Growth performance, carcass and hematological characteristics of rabbits fed graded levels of tiger nuts ( Cyperus esculentus ) ... (p>0.05) difference between treatments. Results demonstrated that (Cyperus esculentus) could be used up to 5% in rabbit's diets without adverse effect on the animals' performance and health.

  5. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  6. Comparison of Ontario Hydro's performance with world power reactors - 1981

    International Nuclear Information System (INIS)

    Dumka, B.R.

    1982-04-01

    The performance of Ontario Hydro's CANDU reactors in 1981 is compared with that of 123 world nuclear power reactors rated at 500 MW(e) or greater. The report is based on data extracted from publications, as well as correspondence with a number of utilities. The basis used is the gross capacity factor, which is defined as gross unit generation divided by the perfect gross output for the period of interest. The lowest of the published turbine and generator design ratings is used to determine the perfect gross output, unless the unit has been proven capable of consistently exceeding this value. The first six reactors in the rankings were CANDU reactors operated by Ontario Hydro

  7. Thermal performances of an insulating structure for a reactor vessel

    International Nuclear Information System (INIS)

    Aranovitch, E.; Crutzen, S.; Le Det, M.; Denis, R.

    1974-12-01

    This report describes the thermal and technological tests performed on a multilayer thermal insulation system for high temperature gas reactors. It includes the description of test facilities, global tests, interpretation of data, and technological tests. Results obtained make it possible to predetermine with a satisfactory precision thermal performances under various nominal conditions

  8. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  9. Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

    International Nuclear Information System (INIS)

    Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

    2014-01-01

    Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

  10. Development of the evaluation methods in reactor safety analyses and core characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  11. Improved performance of parallel surface/packed-bed discharge reactor for indoor VOCs decomposition: optimization of the reactor structure

    International Nuclear Information System (INIS)

    Jiang, Nan; Hui, Chun-Xue; Li, Jie; Lu, Na; Shang, Ke-Feng; Wu, Yan; Mizuno, Akira

    2015-01-01

    The purpose of this paper is to develop a high-efficiency air-cleaning system for volatile organic compounds (VOCs) existing in the workshop of a chemical factory. A novel parallel surface/packed-bed discharge (PSPBD) reactor, which utilized a combination of surface discharge (SD) plasma with packed-bed discharge (PBD) plasma, was designed and employed for VOCs removal in a closed vessel. In order to optimize the structure of the PSPBD reactor, the discharge characteristic, benzene removal efficiency, and energy yield were compared for different discharge lengths, quartz tube diameters, shapes of external high-voltage electrode, packed-bed discharge gaps, and packing pellet sizes, respectively. In the circulation test, 52.8% of benzene was removed and the energy yield achieved 0.79 mg kJ −1 after a 210 min discharge treatment in the PSPBD reactor, which was 10.3% and 0.18 mg kJ −1 higher, respectively, than in the SD reactor, 21.8% and 0.34 mg kJ −1 higher, respectively, than in the PBD reactor at 53 J l −1 . The improved performance in benzene removal and energy yield can be attributed to the plasma chemistry effect of the sequential processing in the PSPBD reactor. The VOCs mineralization and organic intermediates generated during discharge treatment were followed by CO x selectivity and FT-IR analyses. The experimental results indicate that the PSPBD plasma process is an effective and energy-efficient approach for VOCs removal in an indoor environment. (paper)

  12. Gas-liquid reactor / separator: dynamics and operability characteristics

    NARCIS (Netherlands)

    Ranade, V.; Kuipers, J.A.M.; Versteeg, Geert

    1999-01-01

    A comprehensive mathematical model is developed to simulate gas¿liquid reactor in which both, reactants as well as products enter or leave the reactor in gas phase while the reactions take place in liquid phase. A case of first-order reaction (isothermal) was investigated in detail using the dynamic

  13. Potential market and characteristics of low-temperature reactors

    International Nuclear Information System (INIS)

    Lerouge, B.

    1975-01-01

    The low-temperature (100 to 200 deg C) heat market for industrial applications and district heating is very important. Two main studies have been developed: a swimming pool reactor delivering water at 110 deg C and a prestressed concrete vessel reactor delivering water at 200 deg C [fr

  14. Mercury adsorption characteristics of HBr-modified fly ash in an entrained-flow reactor.

    Science.gov (United States)

    Zhang, Yongsheng; Zhao, Lilin; Guo, Ruitao; Song, Na; Wang, Jiawei; Cao, Yan; Orndorff, William; Pan, Wei-ping

    2015-07-01

    In this study, the mercury adsorption characteristics of HBr-modified fly ash in an entrained-flow reactor were investigated through thermal decomposition methods. The results show that the mercury adsorption performance of the HBr-modified fly ash was enhanced significantly. The mercury species adsorbed by unmodified fly ash were HgCl2, HgS and HgO. The mercury adsorbed by HBr-modified fly ash, in the entrained-flow reactor, existed in two forms, HgBr2 and HgO, and the HBr was the dominant factor promoting oxidation of elemental mercury in the entrained-flow reactor. In the current study, the concentration of HgBr2 and HgO in ash from the fine ash vessel was 4.6 times greater than for ash from the coarse ash vessel. The fine ash had better mercury adsorption performance than coarse ash, which is most likely due to the higher specific surface area and longer residence time. Copyright © 2015. Published by Elsevier B.V.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  16. Study on unstable fracture characteristics of light water reactor piping

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1998-08-01

    Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs

  17. Modern control technology for improved nuclear reactor performance

    International Nuclear Information System (INIS)

    Oakes, L.C.

    1986-01-01

    One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level

  18. Characteristics of Flameless Combustion in 3D Highly Porous Reactors under Diesel Injection Conditions

    Directory of Open Access Journals (Sweden)

    M. Weclas

    2013-01-01

    Full Text Available The heat release process in a free volume combustion chamber and in porous reactors has been analyzed under Diesel engine-like conditions. The process has been investigated in a wide range of initial pressures and temperatures simulating engine conditions at the moment when fuel injection starts. The resulting pressure history in both porous reactors and in free volumes significantly depends on the initial pressure and temperature. At lower initial temperatures, the process in porous reactors is accelerated. Combustion in a porous reactor is characterized by heat accumulation in the solid phase of the porous structure and results in reduced pressure peaks and lowered combustion temperature. This depends on reactor heat capacity, pore density, specific surface area, pore structure, and heat transport properties. Characteristic modes of a heat release process in a two-dimensional field of initial pressure and temperature have been selected. There are three characteristic regions represented by a single- and multistep oxidation process (with two or three slopes in the reaction curve and characteristic delay time distribution has been selected in five characteristic ranges. There is a clear qualitative similarity of characteristic modes of the heat release process in a free volume and in porous reactors. A quantitative influence of porous reactor features (heat capacity, pore density, pore structure, specific surface area, and fuel distribution in the reactor volume has been clearly indicated.

  19. Thermic diode performance characteristics and design manual

    Science.gov (United States)

    Bernard, D. E.; Buckley, S.

    1979-01-01

    Thermic diode solar panels are a passive method of space and hot water heating using the thermosyphon principle. Simplified methods of sizing and performing economic analyses of solar heating systems had until now been limited to passive systems. A mathematical model of the thermic diode including its high level of stratification has been constructed allowing its performance characteristics to be studied. Further analysis resulted in a thermic diode design manual based on the f-chart method.

  20. Haematological characteristics and performance of West African ...

    African Journals Online (AJOL)

    The effects of feeding crude petroleum contaminated forage on haematological characteristics and performance of 36 young West African Dwarf (WAD) goats was investigated in order to simulate the impact of real crude oil spillage on livestock and game. Graded levels (0.0, 1.5 and 3.0 g per kg forage) of stabilized “Bonny ...

  1. Growth performance, blood parameters and carcass characteristics ...

    African Journals Online (AJOL)

    This study was carried out with one hundred and twenty (120) day-old marshal chicks to investigate the effect of Maxigrain® enzyme supplementation of corn bran based diets on growth performance, carcass characteristics, haematology and serum biochemistry of broilers in an eight weeks experiment. Four experimental ...

  2. Commission of the European Communities: Review of fast reactor activities performed during 1990

    International Nuclear Information System (INIS)

    Balz, W.

    1991-01-01

    In the field of fast reactors the Commission of the European Communities (CEC) is conducting coordination and harmonization activities at the Brussels headquarters and performing research in its Joint Research Center. The Fast Reactor Coordinating Committee (FRCC) is performing coordination and harmonization activities taking account of the collaboration agreements within the European Fast Reactor (EFR) context. Since the EFR collaboration does not involve all Member States of the European Community the FRCC should establish a link between the EFR countries and other countries. The FRCC discussed R and D activities suitable for a concerted action in a community frame. The Committee also discussed actinide transmutation aspects in LMFBRs. The discussions were based on the results of a study sponsored by the CEC to assess the characteristics of a large core (3600 MWth) with variable actinide content (3-15%). The FRCC received regularly reports on results from current R and D programmes, especially from those related to EFR. (author). 2 figs, 2 tabs

  3. Analysis of tandem mirror reactor performance

    International Nuclear Information System (INIS)

    Wu, K.F.; Campbell, R.B.; Peng, Y.K.M.

    1984-11-01

    Parametric studies are performed using a tandem mirror plasma point model to evaluate the wall loading GAMMA and the physics figure of merit, Q (fusion power/injected power). We explore the relationship among several dominant parameters and determine the impact on the plasma performance of electron cyclotron resonance heating in the plug region. These global particle and energy balance studies were carried out under the constraints of magnetohydrodynamic (MHD) equilibrium and stability and constant magnetic flux, assuming a fixed end-cell geometry. We found that the higher the choke coil fields, the higher the Q, wall loading, and fusion power due to the combination of the increased central-cell field B/sub c/ and density n/sub c/ and the reduced central-cell beta β/sub c/. The MHD stability requirement of constant B/sub c/ 2 β/sub c/ causes the reduction in β/sub c/. In addition, a higher value of fusion power can also be obtained, at a fixed central-cell length, by operating at a lower value of B/sub c/ and a higher value of β/sub c/

  4. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  5. An analysis of CDTN performance in the reactors technology area

    International Nuclear Information System (INIS)

    Pinheiro, R.B.

    1985-01-01

    The author makes an analysis of CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) performance in the reactors technology area, showing difficulties and failures, but emphasizing the particular competence and capacity acquired in this area, as for example: the capacity in codes and methods are of neutronic calculations and nuclear projects, experimental thermohydraulic program, tests services in components and the others. (C.M.) [pt

  6. Breeding characteristics analysis of a commercial fast reactor cooled with sodium liquid

    International Nuclear Information System (INIS)

    Kosaka, N.; Shigehiro, A.

    1982-01-01

    The fast reactor breeding characteristics and its safety is analysed. As reference, for a preliminar analysis, the specifications of Super-Phenix, reactor french of 1200 MWe, are used, varying some parameters after aiming to verify its effects on duplication time. (E.G.) [pt

  7. Development of the reactor lithium ampoule device for research of spectral-luminescent characteristics of nuclear-excited plasma

    Energy Technology Data Exchange (ETDEWEB)

    Batyrbekov, E.G. [National Nuclear Center of RK, Kurchatov (Kazakhstan); Gordienko, Yu. N., E-mail: gordienko@nnc.kz [National Nuclear Center of RK, Kurchatov (Kazakhstan); Ponkratov, Yu. V. [National Nuclear Center of RK, Kurchatov (Kazakhstan); Khasenov, M.U. [PI “National Laboratory Astana”, Astana (Kazakhstan); Tazhibayeva, I.L.; Barsukov, N.I.; Kulsartov, T.V.; Zaurbekova, Zh. A.; Tulubayev, Ye. Yu.; Skakov, M.K. [National Nuclear Center of RK, Kurchatov (Kazakhstan)

    2017-04-15

    Highlights: • The development procedure of the ampoule device for experiments with nuclear-excited plasma under neutron irradiation is described. • The methods of nuclear reactions’ energy conversion into the energy of optical radiation of nuclear-excited plasma are presented. • A scheme of reactor experiments, the experimental facility and experimental device to carry out the reactor experiments are considered. - Abstract: This paper describes the development procedure of the reactor ampoule device to perform the experiments on study of spectral luminescence characteristics of nuclear-excited plasma formed by products of {sup 6}Li(n,α){sup 3}H reaction under neutron irradiation at the IVG.1 M research reactor. The methods of nuclear reactions’ energy conversion into the energy of optical radiation of nuclear-excited plasma are presented. A scheme of reactor experiments, the experimental facility and experimental device to carry out the reactor experiments are considered in paper. The designed ampoule device is totally meets the requirements of irradiation experiments on the IVG.1M reactor.

  8. Density dependence of reactor performance with thermal confinement scalings

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1992-03-01

    Energy confinement scalings for the thermal component of the plasma published thus far have a different dependence on plasma density and input power than do scalings for the total plasma energy. With such thermal scalings, reactor performance (measured by Q, the ratio of the fusion power to the sum of the ohmic and auxiliary input powers) worsens with increasing density. This dependence is the opposite of that found using scalings based on the total plasma energy, indicating that reactor operation concepts may need to be altered if this density dependence is confirmed in future research

  9. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  10. Characteristics of a novel nanosecond DBD microplasma reactor for flow applications

    Science.gov (United States)

    Elkholy, A.; Nijdam, S.; van Veldhuizen, E.; Dam, N.; van Oijen, J.; Ebert, U.; de Goey, L. Philip H.

    2018-05-01

    We present a novel microplasma flow reactor using a dielectric barrier discharge (DBD) driven by repetitive nanosecond high-voltage pulses. Our DBD-based geometry can generate a non-thermal plasma discharge at atmospheric pressure and below in a regular pattern of micro-channels. This reactor can work continuously up to about 100 min in air, depending on the pulse repetition rate and operating pressure. We here present the geometry and main characteristics of the reactor. Pulse energies of 1.46 and 1.3 μJ per channel at atmospheric pressure and 50 mbar, respectively, have been determined by time-resolved measurements of current and voltage. Time-resolved optical emission spectroscopy measurements have been performed to calculate the relative species concentrations and temperatures (vibrational and rotational) of the discharge. The effects of the operating pressure and flow velocity on the discharge intensity have been investigated. In addition, the effective reduced electric field strength {(E/N)}eff} has been obtained from the intensity ratio of vibronic emission bands of molecular nitrogen at different operating pressures and different locations. The derived {(E/N)}eff} increases gradually from about 550 to 4600 Td when decreasing the pressure from 1 bar to 100 mbar. Below 100 mbar, further pressure reduction results in a significant increase in {(E/N)}eff} up to about 10000 Td at 50 mbar.

  11. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  12. Thermohydraulic characteristics under some transient conditions of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Khang, Ngo Phu; An, Tran Khac; Nghiem, Huynh Ton [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Some experimental and theoretical thermal hydraulic characteristics of the Dalat Nuclear Research Reactor are presented, together with some general assessments, from a thermal hydraulic point of view, of its safety under transient conditions. (author). 3 refs., 9 figs., 7 tabs.

  13. Small reactor technical and design characteristics proposed for Indonesia

    International Nuclear Information System (INIS)

    Nurdin, M.

    1992-01-01

    A Team for Small Nuclear Electricity Reactor has been formed in Indonesia since June 1990. It is responsible for assessment and design of a small reactor for electricity and/or sea-water desalination. This concept may become a good alternative for power-plants for small islands and for isolated areas in Indonesia, the system should function economically and environmentally sound. In addition to existing concepts, this presentation deals with modifications proposed in improving reliability and safety of reactor operation. For the size of 200 MWth or more (80 MWe or more), the possibility of designing an internal auxiliary heat removal system is discussed, hence there are two separate heat sinks for the core. Future development works for this concept should be directed in expanding their spectrum of utilization and their contribution to the national energy needs. (author). 7 refs., 4 tabs

  14. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

  15. [Rapid startup and nitrogen removal characteristic of anaerobic ammonium oxidation reactor in packed bed biofilm reactor with suspended carrier].

    Science.gov (United States)

    Chen, Sheng; Sun, De-zhi; Yu, Guang-lu

    2010-03-01

    Packed bed biofilm reactor with suspended carrier was used to cultivate ANAMMOX bacteria with sludge inoculums from WWTP secondary settler. The startup of ANAMMOX reactor was comparatively studied using high nitrogen loading method and low nitrogen loading method with aerobically biofilmed on the carrier, and the nitrogen removal characteristic was further investigated. The results showed that the reactor could be started up successfully within 90 days using low nitrogen loading method, the removal efficiencies of ammonium and nitrite were nearly 100% and the TN removal efficiencywas over 75% , however, the high nitrogen loading method was proved unsuccessfully for startup of ANAMMOX reactor probably because of the inhibition effect of high concentration of ammonium and nitrite. The pH value of effluent was slightly higher than the influent and the pH value can be used as an indicator for the process of ANAMMOX reaction. The packed bed ANAMMOX reactor with suspended carrier showed good characteristics of high nitrogen loading and high removal efficiency, 100% of removal efficiency could be achieved when the influent ammonium and nitrite concentration was lower than 800 mg/L.

  16. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  17. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  18. Performances on nuclear activation analysis by TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Capannesi, G.; Rosada, A.

    1986-01-01

    Progresses in methodological research and connected applications in the field of activation analysis are introduced. Some peculiar characteristics on the TRIGA MARK II reactor have enabled the possibility of obtaining interesting results. The particular, the rotating radiation device Lazy Susan, with a capability of 40 positionings, permits homogeneity in neutron flux and energy spectrum stability within 15%. High level of precision and accuracy are obtained in analytic. Applications of major interest have been: - reference material certification; - forensic applications; - electrolytic cell productivity evaluation. The TRIGA MARK II reactor is equipped with a thermal column throughout a D 2 O diaphragm with a thickness of 70 cm. The available neutron flux has no fast and epithermal components. Via this facility a method has been tested for the instrumental determination of Al in Si metal of solar and electronic degree. (author)

  19. Core performance of equilibrium fast reactors for different coolant materials and fuel types

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

    Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

  20. Influence of remaining fission products in low-decontaminated fuel on reactor core characteristics

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2002-07-01

    Design study of core, fuel and related fuel cycle system with low-decontaminated fuel has been performed in the framework of the feasibility study (F/S) on commercialized fast reactor cycle systems. This report summarizes the influence on core characteristics of remaining fission products (FPs) in low-decontaminated fuel related to the reprocessing systems nominated in F/S phase I. For simple treatment of the remaining FPs in core neutronics calculation the representative nuclide method parameterized by the FP equivalent coefficient and the FP volume fraction was developed, which enabled an efficient evaluation procedure. As a result of the investigation on the sodium cooled fast reactor with MOX fuel designed in fiscal year 1999, it was found that the pyrochemical reprocessing with molten salt (the RIAR method) brought the largest influence. Nevertheless, it was still within the allowable range. Assuming an infinite-times recycling, the alternations in core characteristics were evaluated as follows: increment of burnup reactivity by 0.5%Δk/kk', decrement of breeding ratio by 0.04, increment of sodium void reactivity by 0.1x10 -2 Δk/kk' and decrement of Doppler constant (in absolute value) by 0.7x10 -3 Tdk/dT. (author)

  1. Hydraulic characteristics of a fast reactor fuel subassembly: An experimental investigation

    International Nuclear Information System (INIS)

    Padmakumar, G.; Velusamy, K.; Prasad, B.V.S.S.; Rajan, K.K.

    2017-01-01

    Highlights: • Fuel subassembly bundle geometry is studied for its hydraulic behaviour. • The results are also compared with data available in literature. • All flow regimes viz. laminar, transition and turbulent is covered for the study. • Pressure drop across different regions of subassembly was also determined. • The effect of external blockage is also studied and reported. - Abstract: Fuel subassemblies of a fast reactor consist of fuel pin bundle with helically wound spacer wires, arranged in a triangular pitch within a hexagonal wrapper. The fuel pins are located within the subassembly. Further the subassembly comprises of a diffuser where the cross section changes from cylindrical to hexagonal, mixing plenum before the exit of pin bundle and a specially designed blockage adapter. Accurate assessment of the pressure drop in the fuel subassembly is essential to ensure adequate core cooling and design of sodium pump. Experimental determination of pressure drop characteristics in the subassembly by simulating the hydraulic condition in the subassemblies of the reactor core is considered essential as a better choice as correlations reported in the literature cannot be directly used for all the complex regions present in the subassembly. This is due to the fact that flows in the interconnecting sections are highly under developed. Further, the flow regime in a fuel subassembly varies from laminar (during shutdown heat removal under natural convection) to completely turbulent under full power condition. To understand the hydraulic characteristics of the 500 MWe Proto type Fast Breeder Reactor (PFBR) fuel subassembly, an experimental facility has been commissioned. Experiments on full scale subassembly with dummy fuel pins have been performed using water as simulant. Experiments have been conducted covering a wide range of Reynolds number encompassing laminar, transition and turbulent regimes. In the rod bundle, no abrupt changes in friction factor were

  2. Scram characteristics of the control rods of a pressurized water reactor under seismic conditions

    International Nuclear Information System (INIS)

    Fujita, Katsuhisa; Shinohara, Yoshikazu; Nakatogawa, Tetsuto; Nanbu, Kiyoshi; Nomura, Tomonori.

    1987-01-01

    Control rod drop verification experiments of a pressurized water reactor under seismic conditions are performed to confirm the insertion function of control rods into a core. To evaluate these tests, computer simulations are performed. A fuel assembly, control rods, guide tube and other associated structures are immersed in a water tank, and shaken by four hydraulic shakers. The scram time of control rods under seismic conditions was measured, and confirmed to meet the scram function. Moreover, vibrational response characteristics of core structures and dropping behavior of control rods in consideration of collisions are calculated by using a finite difference method. The behavior of the dropping control rods and the scram time obtained by the computer simulation show a very good agreement with the verification experimental results. (author)

  3. WWER reactor fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Stefanova, S.; Chantoin, P.; Kolev, I.

    1994-01-01

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

  4. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  5. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  6. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  7. WWER reactor fuel performance, modelling and experimental support. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Chantoin, P; Kolev, I [eds.

    1994-12-31

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

  8. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  9. Operational limitations of light water reactors relating to fuel performance

    International Nuclear Information System (INIS)

    Cheng, H.S.

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

  10. Study of heat exchange characteristics of the Dalat Nuclear Reactor

    International Nuclear Information System (INIS)

    An, N.K.; Huy, N.Q.

    1989-01-01

    This report is presented some experimental data and related theoretical computations concerning the thermal exchange system under normal operating or accidental conditions from the thermodynamic point of view. In the normal operation, the reactor operates under safety condition T max fuel=96.2 degree C. Under LOFA condition, the heat exchage process is still realized, therefore, we should determine the allowable limits of the thermal regime at power and at shut down condition

  11. Neutron radiation characteristics of the IVth generation reactor spent fuel

    Science.gov (United States)

    Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

    2018-03-01

    Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

  12. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  13. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  14. Summary report of the experimental fast reactor JOYO MK-III performance test

    International Nuclear Information System (INIS)

    Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

    2004-03-01

    An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

  15. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  16. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  17. Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

    Science.gov (United States)

    Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

  18. Characteristics and performances of electronic personal dosemeters

    International Nuclear Information System (INIS)

    Aubert, B.

    2002-01-01

    The regulations have made obligation for 2 years to measure and analyse the amounts of radiations actually received during an operation. The whole of these measurements taken uninterrupted for an immediate reading is indicated like the operational dosimetry, which is carried out with the means of personal electronic dosemeters. This study analyses the legislation relating to this type of dosimetry as well as the requirements in medical environment, and presents an assessment of the characteristics and performances of the devices available on the French market at the beginning of 2002 starting from the information provided by the various manufacturers. (author)

  19. Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

    International Nuclear Information System (INIS)

    Kim, Taek Kyum; Downar, Thomas J.

    2002-01-01

    Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

  20. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

    The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

  1. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  2. Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

    Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

  3. Theoretical and experimental analysis of fast reactor fuel performance

    International Nuclear Information System (INIS)

    Kummerer, K.R.; Freund, D.; Steiner, H.

    1982-09-01

    In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

  4. Impact of reactor water chemistry on cladding performance

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  5. Overcoming the effects of stress on reactor operator performance

    International Nuclear Information System (INIS)

    He Xuhong; Wei Li; Zhao Bingquan

    2003-01-01

    Reactor operators may be exposed to significant levels of stress during plant emergencies and their performance may be affected by the stress. This paper first identified the potential sources of stress in the nuclear power plant, then discussed the ways in which stress is likely to affect the reactor operators, and finally identified several training approaches for reducing or eliminating stress effects. The challenges for effective stress reducing training may seem daunting, yet the challenges are real and must be addressed. This paper reviewed researches in training design, knowledge and skill acquisition, and training transfer point to a number of strategies that can be used to address these challenges and lead to more effective training and development. (author)

  6. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  7. Value addition initiatives for CANDU reactor operation performance

    International Nuclear Information System (INIS)

    Chugh, V.; Parmar, R.; Schut, J.; Sherin, J.; Xie, H.; Zobin, D.

    2013-01-01

    Recently, AMEC NSS initiated projects for CANDU® station performance engineering with potentially high returns for the utilities. This paper discusses three initiatives. Firstly, optimization of instrument calibration interval from 1 to 3 years will reduce time commitments on the maintenance resources on top of financial savings ~$3,500 per instrument. Secondly, reactor thermal power uncertainty assessment shows the level of operation which is believed to have an over-conservative margin that can be used to increase power by up to 0.75%. Finally, as an alternative means for controlling Reactor Inlet Header Temperature (RIHT), physical modifications to the High Pressure (HP) feedwater heaters can be useful for partially recovering RIHT resulting in increased production by 10-12 MWe. (author)

  8. Thermal performance and efficiency of supercritical nuclear reactors

    International Nuclear Information System (INIS)

    Romney Duffey; Tracy Zhou; Hussam Khartabil

    2009-01-01

    The paper reviews the major advances and innovative aspects of the thermal performance of recent concepts for super-critical water-cooled nuclear reactors (SCWR). The concepts are based on the extensive experience in the thermal power industry with super and ultra-supercritical boilers and turbines. The challenges and goals of increased efficiency, reduced cost, enhanced safety and co-generation have been pursued over the last ten years, and have resulted both in viable concepts and a vibrant defined R and D effort. The supercritical concept has wide acceptance among industry, as it reflects standard engineering practices and current thermal plant technology that is being already deployed. The SCWR concept represents a continuous development of water-cooled reactor technology, which utilizes the best and latest advances made in the thermal power industry. (author)

  9. Performance of rotary kiln reactor for the elephant grass pyrolysis.

    Science.gov (United States)

    De Conto, D; Silvestre, W P; Baldasso, C; Godinho, M

    2016-10-01

    The influence of process conditions (rotary speed/temperature) on the performance of a rotary kiln reactor for non-catalytic pyrolysis of a perennial grass (elephant grass) was investigated. The product yields, the production of non-condensable gases as well as the biochar properties were evaluated. The maximum H2 yield was close to that observed for catalytic pyrolysis processes, while the bio-oil yield was higher than reported for pyrolysis of other biomass in rotary kiln reactors. A H2/CO ratio suitable for Fischer-Tropsch synthesis (FTS) was obtained. The biochars presented an alkaline pH (above 10) and interesting contents of nutrients, as well as low electrical conductivity, indicating a high potential as soil amendment. Copyright © 2016 Elsevier Ltd. All rights reserved.

  10. Overcoming the effects of stress on reactor operator performance

    Energy Technology Data Exchange (ETDEWEB)

    He Xuhong; Wei Li; Zhao Bingquan [Tsinghua Univ., Nuclear Power Plant Simulation Training Center, Beijing (China)

    2003-03-01

    Reactor operators may be exposed to significant levels of stress during plant emergencies and their performance may be affected by the stress. This paper first identified the potential sources of stress in the nuclear power plant, then discussed the ways in which stress is likely to affect the reactor operators, and finally identified several training approaches for reducing or eliminating stress effects. The challenges for effective stress reducing training may seem daunting, yet the challenges are real and must be addressed. This paper reviewed researches in training design, knowledge and skill acquisition, and training transfer point to a number of strategies that can be used to address these challenges and lead to more effective training and development. (author)

  11. Characteristics explaining performance in downhill mountain biking.

    Science.gov (United States)

    Chidley, Joel B; MacGregor, Alexandra L; Martin, Caoimhe; Arthur, Calum A; Macdonald, Jamie H

    2015-03-01

    To identify physiological, psychological, and skill characteristics that explain performance in downhill (DH) mountain-bike racing. Four studies were used to (1) identify factors potentially contributing to DH performance (using an expert focus group), (2) develop and validate a measure of rider skill (using video analysis and expert judge evaluation), (3) evaluate whether physiological, psychological, and skill variables contribute to performance at a DH competition, and (4) test the specific contribution of aerobic capacity to DH performance. STUDY 1 identified aerobic capacity, handgrip endurance, anaerobic power, rider skill, and self-confidence as potentially important for DH. In study 2 the rider-skill measure displayed good interrater reliability. Study 3 found that rider skill and handgrip endurance were significantly related to DH ride time (β=-0.76 and -0.14, respectively; R2=.73), with exploratory analyses suggesting that DH ride time may also be influenced by self-confidence and aerobic capacity. Study 4 confirmed aerobic capacity as an important variable influencing DH performance (for a DH ride, mean oxygen uptake was 49±5 mL·kg(-1)·min(-1), and 90% of the ride was completed above the 1st ventilatory threshold). In order of importance, rider skill, handgrip endurance, self-confidence, and aerobic capacity were identified as variables influencing DH performance. Practically, this study provides a novel assessment of rider skill that could be used by coaches to monitor training and identify talent. Novel intervention targets to enhance DH performance were also identified, including self-confidence and aerobic capacity.

  12. Characteristics of isotope-selective chemical reactor with gas-separating device

    International Nuclear Information System (INIS)

    Gorshunov, N.M.; Kalitin, S.A.; Laguntsov, N.I.; Neshchimenko, Yu.P.; Sulaberidze, G.A.

    1988-01-01

    A study was made on characteristics of separating stage, composed of isotope-selective chemical (or photochemical) reactor and membrane separating cascade (MSC), designated for separation of isotope-enriched products from lean reagents. MSC represents the counterflow cascade for separation of two-component mixtures. Calculations show that for the process of carton isotope separation the electric power expences for MSC operation are equal to 20 kWxh/g of CO 2 final product at 13 C isotope content in it equal to 75%. Application of the membrane gas-separating cascade at rather small electric power expenses enables to perform cascading of isotope separation in the course of nonequilibrium chemical reactions

  13. Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

    1987-01-01

    Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (author)

  14. ELFR: The European Lead Fast Reactor. Design, Safety Approach and Safety Characteristics

    International Nuclear Information System (INIS)

    Alemberti, Alessandro

    2012-01-01

    • In the framework of the LEADER project, the safety approach for a Lead cooled fast reactor has been defined and, in particular, all the possible challenges to the main safety functions and their mechanisms have been specified, in order to better define the needed provisions. • On the basis of the above and taking into account the results of the safety analyses performed during previous project (ELSY), a reference configuration of the ELFR plant has been consolidated, by improving and updating the plant design features. In particular, the emerged safety concerns have been analyzed in the LEADER project and a new set of design options and safety provisions have been proposed. • The combination of favourable Lead coolant inherent characteristics and plant design features, specifically developed to face identified challenges, resulted in a very robust and forgiving design, even in very extreme conditions, as a Fukushima-like scenario

  15. Influence of core model parameters on the characteristics of neutron beams of the research reactor

    Directory of Open Access Journals (Sweden)

    N. A. Khafizova

    2013-12-01

    Full Text Available IRT MEPhI reactor is equipped with a number of facilities at horizontal experimental channels (HEC. Knowing of parameters influencing spatio-angular distribution of irradiation fields is essential for each application area. The research for neutron capture therapy (NCT facility at HEC of the reactor was made. Calculation methods have been used to estimate how the reactor core parameters influence neutron beam characteristics at the HEC output. The impact of neutron source model in Monte Carlo calculations by MCNP code on the parameters of neutron and secondary photon field at the output of irradiation beam tubes of research reactor is estimated. The study shows that specifying neutron source with fission reaction rate distribution in SDEF option gives almost the same results as criticality calculation considered the most accurate. Our calculations show that changes of the core operational parameters have insignificant influence on characteristics of neutron beams at HEC output.

  16. Steady state characteristics of acclimated hydrogenotrophic methanogens on inorganic substrate in continuous chemostat reactors.

    Science.gov (United States)

    Ako, Olga Y; Kitamura, Y; Intabon, K; Satake, T

    2008-09-01

    A Monod model has been used to describe the steady state characteristics of the acclimated mesophilic hydrogenotrophic methanogens in experimental chemostat reactors. The bacteria were fed with mineral salts and specific trace metals and a H(2)/CO(2) supply was used as a single limited substrate. Under steady state conditions, the growth yield (Y(CH4)) reached 11.66 g cells per mmol of H(2)/CO(2) consumed. The daily cells generation average was 5.67 x 10(11), 5.25 x 10(11), 4.2 x 10(11) and 2.1 x 10(11) cells/l-culture for the dilutions 0.071/d, 0.083/d, 0.1/d and 0.125/d, respectively. The maximum specific growth rate (mu(max)) and the Monod half-saturation coefficient (K(S)) were 0.15/d and 0.82 g/L, respectively. Using these results, the reactor performance was simulated. During the steady state, the simulation predicts the dependence of the H(2)/CO(2) concentration (S) and the cell concentration (X) on the dilution rate. The model fitted the experimental data well and was able to yield a maximum methanogenic activity of 0.24 L CH(4)/g VSS.d. The dilution rate was estimated to be 0.1/d. At the dilution rate of 0.14/d, the exponential cells washout was achieved.

  17. Aerobic granules formation and nutrients removal characteristics in sequencing batch airlift reactor (SBAR) at low temperature

    International Nuclear Information System (INIS)

    Bao Ruiling; Yu Shuili; Shi Wenxin; Zhang Xuedong; Wang Yulan

    2009-01-01

    To understand the effect of low temperature on the formation of aerobic granules and their nutrient removal characteristics, an aerobic granular sequencing batch airlift reactor (SBAR) has been operated at 10 deg. C using a mixed carbon source of glucose and sodium acetate. The results showed that aerobic granules were obtained and that the reactor performed in stable manner under the applied conditions. The granules had a compact structure and a clear out-surface. The average parameters of the granules were: diameter 3.4 mm, wet density 1.036 g mL -1 , sludge volume index 37 mL g -1 , and settling velocity 18.6-65.1 cm min -1 . Nitrite accumulation was observed, with a nitrite accumulation rate (NO 2 - -N/NO x - -N) between 35% and 43% at the beginning of the start-up stage. During the stable stage, NO x was present at a level below the detection limit. However, when the influent COD concentration was halved (resulting in COD/N a reduction of the COD/N from 20:1 to 10:1) nitrite accumulation was observed once more with an effluent nitrite accumulation rate of 94.8%. Phosphorus release was observed in the static feeding phase and also during the initial 20-30 min of the aerobic phase. Neither the low temperature nor adjustment of the COD/P ratio from 100:1 to 25:1 had any influence on the phosphorus removal efficiency under the operating conditions. In the granular reactor with the influent load rates for COD, NH 4 + -N, and PO 4 3- -P of 1.2-2.4, 0.112 and 0.012-0.024 kg m -3 d -1 , the respective removal efficiencies at low temperature were 90.6-95.4%, 72.8-82.1% and 95.8-97.9%.

  18. Performance Variation of Spent Resin in Mixed Bed From Water Purifying System of Xi'an Pulse Reactor

    International Nuclear Information System (INIS)

    Li Hua; Ma Yan; Xiao Yan; Liu Yueheng; Yang Yongqing

    2010-01-01

    Detailed physical and chemical characteristic analysis was performed on the spent cation and anion resins in the mixed bed from Xi'an Pulse Reactor water purifying system.The exchange performance variations of used resins and the contributions from different factors to the variation were discussed.Based on the obtained information of the impurities in the used resin, the contamination state of the water in the Xi'an Pulse Reactor water pool, the corrosion state of the structural material in the reactor was presented. The spent anion resin almost completely losses its exchange performance,while the remaining exchange capacity in the spent cation resin is still high.The radiation field from the reactor operation contributes little to the degradation of the performance of the resins. The exchange capacity loss of the spent anion resin is due to the exchange of its active groups into abundant carbonate and a certain amount of organics. The impurity amount in the anion and cation exchange resins is low,which suggests(that) the water in the Xi'an Pulse Reactor water pool is little contaminated. A certain extent of corrosion is occurred on the structural material in the swimming pool of the reactor. The results provide important referential data for the operational safety of the water purifying system of similar research reactor. (authors)

  19. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    International Nuclear Information System (INIS)

    Hu Jian; Jiang Nan; Li Jie; Shang Kefeng; Lu Na; Wu Yan; Mizuno Akira

    2016-01-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. (paper)

  20. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    Science.gov (United States)

    Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

    2016-03-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

  1. Analysis for RSG-GAS operational characteristics of reactor cooling system

    International Nuclear Information System (INIS)

    Nurhappy, T.

    1998-01-01

    Analysis of operational characteristics of reactor cooling systems (JE01 and PA) is aimed at determining the effects of operation and maintenance patterns to the operational characteristic of the system. Analysis is carried out by virtue of the operating and maintenance data from 1987 to 1997, comprising the operating hours (duration) and data on operating failures of the systems. Results of study show that, either separately or jointly, the operating and maintenance patterns will qualitatively affect the operational characteristic of the systems

  2. Performance of a UASB reactor treating coffee wet wastewater

    International Nuclear Information System (INIS)

    Guardia Puebla, Yans; Rodríguez Pérez, Suyén; Janet Jiménez Hernández; Sánchez Girón, Víctor

    2014-01-01

    The present work shows the results obtained in the anaerobic digestion process of coffee wet wastewater processing. An UASB anaerobic reactor was operated in single-stage in mesophilic temperature controlled conditions (37±1ºC). The effect of both organic loading rate (OLR) and hydraulic retention time (HRT) in the anaerobic digestion of coffee wet wastewater was investigated. The OLR values considered in the single-stage UASB reactor varied in a range of 3,6-4,1 kgCOD m-3 d-1 and the HRT stayed in a range of 21,5-15,5 hours. The evaluation results show that the best performance of UASB reactor in single-stage was obtained at OLR of 3,6 kg COD m-3 d-1 with an average value of total and soluble COD removal of 77,2% and 83,4%, respectively, and average methane concentration in biogas of 61%. The present study suggests that the anaerobic digestion is suitable to treating coffee wet wastewater. (author)

  3. Analysis on flow characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin

    1997-06-01

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)

  4. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    International Nuclear Information System (INIS)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K.; Kim, J. H.

    2015-01-01

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed

  5. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K. [Changwon National University, Changwon (Korea, Republic of); Kim, J. H. [Daejeon University, Daejeon (Korea, Republic of)

    2015-03-15

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed.

  6. Study and choice of main characteristics of fast reactor - Effective minor actinide burner

    International Nuclear Information System (INIS)

    Krivitski, I.Yu.; Matveev, V.I.; Poplavski, V.M.

    1996-01-01

    This paper presents the principal design and performance data of advanced fast power reactor core for plutonium and actinides burning. Some information concerning the Russian programme of plutonium utilization are also presented. (author). 2 refs, 4 figs, 5 tabs

  7. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  8. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

  9. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-01-01

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

  10. Neutronic characteristics of linear-assembly breed-and-burn reactors

    International Nuclear Information System (INIS)

    Petroski, Robert; Forget, Benoit; Forsberg, Charles

    2012-01-01

    Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

  11. Microbiology and performance of a methanogenic biofilm reactor during the start-up period.

    Science.gov (United States)

    Cresson, R; Dabert, P; Bernet, N

    2009-03-01

    To understand the interactions between anaerobic biofilm development and process performances during the start-up period of methanogenic biofilm reactor. Two methanogenic inverse turbulent bed reactors have been started and monitored for 81 days. Biofilm development (adhesion, growth, population dynamic) and characteristics (biodiversity, structure) were investigated using molecular tools (PCR-SSCP, FISH-CSLM). Identification of the dominant populations, in relation to process performances and to the present knowledge of their metabolic activities, was used to propose a global scheme of the degradation routes involved. The inoculum, which determines the microbial species present in the biofilm influences bioreactor performances during the start-up period. FISH observations revealed a homogeneous distribution of the Archaea and bacterial populations inside the biofilm. This study points out the link between biodiversity, functional stability and methanogenic process performances during start-up of anaerobic biofilm reactor. It shows that inoculum and substrate composition greatly influence biodiversity, physiology and structure of the biofilm. The combination of molecular techniques associated to a biochemical engineering approach is useful to get relevant information on the microbiology of a methanogenic growing biofilm, in relation with the start-up of the process.

  12. Rotating bed reactor for CLC: Bed characteristics dependencies on internal gas mixing

    International Nuclear Information System (INIS)

    Håkonsen, Silje Fosse; Grande, Carlos A.; Blom, Richard

    2014-01-01

    Highlights: • A mathematical model for the rotating CLC reactor has been developed. • The model reflects the gas distribution in the reactor during CLC operation. • Radial dispersion in the rotating bed is the main cause for internal gas mixing. • The model can be used to optimize the reactor design and particle characteristics. - Abstract: A newly designed continuous lab-scale rotating bed reactor for chemical looping combustion using CuO/Al 2 O 3 oxygen carrier spheres and methane as fuel gives around 90% CH 4 conversion and >90% CO 2 capture efficiency based on converted methane at 800 °C. However, from a series of experiments using a broad range of operating conditions potential CO 2 purities only in the range 20–65% were yielded, mostly due to nitrogen slip from the air side of the reactor into the effluent CO 2 stream. A mathematical model was developed intending to understand the air-mixing phenomena. The model clearly reflects the gas slippage tendencies observed when varying the process conditions such as rotation frequency, gas flow and the flow if inert gas in the two sectors dividing the air and fuel side of the reactor. Based on the results, it is believed that significant improvements can be made to reduce gas mixing in future modified and scaled-up reactor versions

  13. Fluid Flow Characteristic Simulation of the Original TRIGA 2000 Reactor Design Using Computational Fluid Dynamics Code

    International Nuclear Information System (INIS)

    Fiantini, Rosalina; Umar, Efrizon

    2010-01-01

    Common energy crisis has modified the national energy policy which is in the beginning based on natural resources becoming based on technology, therefore the capability to understanding the basic and applied science is needed to supporting those policies. National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. The previous analysis of the modification TRIGA 2000 Reactor design indicates that forced convection of the primary coolant system put on an effect to the flow characteristic in the reactor core, but relatively insignificant effect to the flow velocity in the reactor core. In this analysis, the lid of reactor core is closed. However the forced convection effect is still presented. This analysis shows the fluid flow velocity vector in the model area without exception. Result of this analysis indicates that in the original design of TRIGA 2000 reactor, there is still forced convection effects occur but less than in the modified TRIGA 2000 design.

  14. Operational characteristics of the CALIBAN fast pulse reactor

    International Nuclear Information System (INIS)

    Cortella, J.; Reberol, R.; Vanel, M.

    1976-01-01

    CALIBAN is a FPR operated at CEA-Valduc Center since 1971. It has been designed as a fast neutron irradiation source and its environment is specific for this utilization. To date, it delivered more than 400 bursts and the fuel integrated about 5.10 19 fissions. The main characteristics are: - cylindirical core 113 kg U 235 - Mo 10% alloy - integrated dose in a burst in the central 2.5cm diam cavity:3.10 14 n.cm -2 - integrated dose in a burst outside of the core:5.10 13 n.cm -2 - pulse width:50μs A special effort was made in measuring the spectrometric and dosimetric neutron and gamma characteristics. Some results will be presented here. (auth.)

  15. Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

    International Nuclear Information System (INIS)

    Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

    2014-01-01

    The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

  16. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  17. Parametric study on thermal-hydraulic characteristics of high conversion light water reactor

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Fujii, Sadao.

    1988-11-01

    To assess the feasibility of high conversion light water reactors (HCLWRs) from the thermal-hydraulic viewpoint, parametric study on thermal-hydraulic characteristics of HCLWR has been carried out by using a unit cell model. It is assumed that a HCLWR core is contained in a current 1000 MWe PWR plant. At the present study, reactor core parameters such as fuel pin diameter, pitch, core height and linear heat rate are widely and parametrically changed to survey the relation between these parameters and the basic thermal-hydraulic characteristics, i.e. maximum fuel temperature, minimum DNBR, reduction of reactor thermal output and so on. The validity of the unit cell model used has been ensured by comparison with the result of a subchannel analysis carried out for a whole core. (author)

  18. Characteristics of D-3He fueled frc reactor: ARTEMIS-L

    International Nuclear Information System (INIS)

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

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

  19. Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

    Science.gov (United States)

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

    1993-11-01

    The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

  20. Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

  1. Blanket materials for fusion reactors: comparisons of thermochemical performance

    International Nuclear Information System (INIS)

    Johnson, C.E.; Fischer, A.K.; Tetenbaum, M.

    1984-01-01

    Thermodynamic calculations have been made to predict the thermochemical performance of the fusion reactor breeder materials, Li 2 O, LiAlO 2 , and Li 4 SiO 4 in the temperature range 900 to 1300 0 K and in the oxygen activity range 10 -25 to 10 -5 . Except for a portion of these ranges, the performance of LiAlO 2 is predicted to be better than that of Li 2 O and Li 4 SiO 4 . The protium purge technique for enhancing tritium release is explored for the Li 2 O system; it appears advantageous at higher temperatures but should be used cautiously at lower temperatures. Oxygen activity is an important variable in these systems and must be considered in executing and interpreting measurements on rates of tritium release, the form of released tritium, diffusion of tritiated species and their identities, retention of tritium in the condensed phase, and solubility of hydrogen isotope gases

  2. Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

    International Nuclear Information System (INIS)

    Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

    2012-01-01

    In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

  3. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    Klein, W.

    1987-01-01

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m -2 . Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN 2 , SiC 2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  4. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  5. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  6. Measurement of transient hydrodynamic characteristics of the reactor RA primary cooling system

    International Nuclear Information System (INIS)

    Jovic, L.; Majstorovic, D.; Zeljkovic, I.

    1987-01-01

    Experimental study of transient hydrodynamic characteristics of the research nuclear reactor RA by simultaneous measurements of fluid flow and pressure on several locations of the RA primary coolant system is done. Loss of electric power transient on the main circulation pumps is simulated. measurement methodology, data processing and results of measured data analysis are given. (author)

  7. Combustion, cofiring and emissions characteristics of torrefied biomass in a drop tube reactor

    International Nuclear Information System (INIS)

    Ndibe, Collins; Maier, Jörg; Scheffknecht, Günter

    2015-01-01

    The study investigates cofiring characteristics of torrefied biomass fuels at 50% thermal shares with coals and 100% combustion cases. Experiments were carried out in a 20 kW, electrically heated, drop-tube reactor. Fuels used include a range of torrefied biomass fuels, non-thermally treated white wood pellets, a high volatile bituminous coal and a lignite coal. The reactor was maintained at 1200 °C while the overall stoichiometric ratio was kept constant at 1.15 for all combustion cases. Measurements were performed to evaluate combustion reactivity, emissions and burn-out. Torrefied biomass fuels in comparison to non-thermally treated wood contain a lower amount of volatiles. For the tests performed at a similar particle size distribution, the reduced volatile content did not impact combustion reactivity significantly. Delay in combustion was only observed for test fuel with a lower amount of fine particles. The particle size distribution of the pulverised grinds therefore impacts combustion reactivity more. Sulphur and nitrogen contents of woody biomass fuels are low. Blending woody biomass with coal lowers the emissions of SO 2 mainly as a result of dilution. NO X emissions have a more complex dependency on the nitrogen content. Factors such as volatile content of the fuels, fuel type, furnace and burner configurations also impact the final NO X emissions. In comparison to unstaged combustion, the nitrogen conversion to NO X declined from 34% to 9% for air-staged co-combustion of torrefied biomass and hard coal. For the air-staged mono-combustion cases, nitrogen conversion to NO X declined from between 42% and 48% to about 10%–14%. - Highlights: • Impact of torrefaction on cofiring was studied at high heating rates in a drop tube. • Cofiring of torrefied biomasses at high thermal shares (50% and higher) is feasible. • Particle size impacts biomass combustion reactivity more than torrefaction. • In a drop tube reactor, torrefaction has no negative

  8. Physical characteristics of non-fuel assembly reactor components

    International Nuclear Information System (INIS)

    Hawkes, E.C.

    1994-09-01

    The primary objective of this report is to enhance the utility of the Characteristics Data Base (CDB). This has been accomplished by providing a pictorial representation of the principal non-fuel assembly (NFA) components along with a tabular summary of key information about each type of component. This report is intended for use as an adjunct to the CDB. Toward this end, the report may be used either as a complement to the detailed descriptions in the CDB, or as a stand-alone document that acts as an illustrated abstract of the CDB. Line drawings of major NFA components are included. Data not provided in the CDB are also included. Summary descriptions of each component are given in tabular format

  9. Performance of an innovative multi-stage anaerobic reactor during ...

    African Journals Online (AJOL)

    Start-up of an anaerobic reactor is a relatively delicate process and depends on various factors such as wastewater composition, available inoculum, operating conditions and reactor configuration. Accordingly, systematized operational procedures are important, mainly during the start-up of an anaerobic reactor.

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

    NARCIS (Netherlands)

    RAY, AK

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

  12. Nanotechnological inventions considerably improve performance characteristics

    Directory of Open Access Journals (Sweden)

    VLASOV Vladimir Alexeevich

    2014-06-01

    Full Text Available The invention «The method of production of carbon nanomaterial (RU 2509053» can be used as an additive for concretes and polymers which significantly improves their performance characteristics. The method of production of carbon nanomaterial consists of the following stages: preliminary preparation of sphagnous moss when it is refined from foreign admixtures, dried up to 10% humidity and ground, then ground material is exposed to pyrolysis under the temperature 850–950оC for 1–2 hours and cooled up to the environment temperature. After that amorphous carbon obtained in pyrolysis is treated with mechanical activation in the variable planetary mill for 7–10 hours. The invention makes it possible to provide increased outcome of nanotubes with high cleanliness. The invention «The method of production of nanodispersed metal powders and alloys of them (RU 2509626» relates to the powder metallurgy. Powder metal chloride or powder mixture at least of two metal chlorides is treated in the environment of the water steam which is supplied in reaction space at the rate of 50–100 ml/min at the temperature 400–800оC at the presence of absorbent carbon or introducing carbon oxide (II obtained during dissolution of formic acid HCOOH. The invention provides reliable production of nanodispersed metal powders and alloys of them from 3-d metal range: Ni, Co, Cu, Fe, Zn which can be used in powder metallurgy to improve baking process, in chemical industry as the fillers of polymers and reaction catalysts; as additives to anticorrosive covers, etc.

  13. Probabilistic assessment of light water reactor fuel performance

    International Nuclear Information System (INIS)

    Misfeldt, I.

    1978-10-01

    A computer system for the statistical evaluation of LWR fuel performance has been developed. The computer code FRP, Fuel Reliability Predictor, calculates the distributions for parameters characterizing the fuel performance and failure probability. The statistical methods employed are either Monte Carlo simulations or low order Taylor approximation. Included in the computer system is a deterministic fuel performance code, which has been verified by comparison with data from irradiation experiments. The distributions for all material data utilized in the fuel simulations are estimations from the best available information in the literature. For the failure prediction, a stress corrosion failure criterion has been derived. The failure criterion is based on data from out-of-reactor stress corrosion experiments performed on unirradiated and irradiated zircaloy with iodine present. By means of an example the typical distributions of the variables characterizing the fuel performance and the accuracy of the methods themselves have been investigated. The application of the computer system is illustrated by a number of examples, these include the evaluation of irradiation experiments, design comparisons, and analyses of minor accidents. (author)

  14. 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...

  15. Fort St. Vrain reactor performance and operation to full power

    International Nuclear Information System (INIS)

    Simon, W.A.; Bramblett, G.C.

    1982-01-01

    The Fort St. Vrain Nuclear Generating Station, powered by a high-temperature gas-cooled reactor (HTGR), has now been tested to full thermal power. Testing was conducted for the dual purposes of demonstrating component and system capability as a part of the rise-to-power program and determining core fluctuation/redistribution behavior under full power conditions. Both objectives were met. Full power performance of all major components and the achievement of nearly all design objectives has been verified. In addition, the tests showed that the fluctuation phenomenon has been corrected. Core region outlet temperature redistributions have been characterized, related to a physical mechanism, and shown to be inconsequential for overall plant operation

  16. Increased performance of continuous stirred tank reactor with calcium supplementation

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Zhuliang; Yang, Haijun; Zhi, Xiaohua; Shen, Jianquan [Beijing National Laboratory for Molecular Sciences (BNLMS), New Materials Laboratory, Institute of Chemistry, Chinese Academy of Sciences, Beijing 100190 (China)

    2010-04-15

    Continuous biohydrogen production with calcium supplementation at low hydraulic retention time (HRT) in a continuous stirred tank reactor (CSTR) was studied to maximize the hydrogen productivity of anaerobic mixed cultures. After stable operations at HRT of 8-4 h, the bioreactor became unstable when the HRT was lowered to 2 h. Supplementation of 100 mg/L calcium at HRT 2 h improved the operation stability through enhancement of cell retention with almost two-fold increase in cell density than that without calcium addition. Hydrogen production rate and hydrogen yield reached 24.5 L/d/L and 3.74 mol H{sub 2}/mol sucrose, respectively, both of which were the highest values our group have ever achieved. The results showed that calcium supplementation can be an effective way to improve the performance of CSTR at low HRT. (author)

  17. Performance of Fragema fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Dumont, A.; Ravier, G.; Ballot, B.

    1986-06-01

    FRAGEMA fuel operating experience in power reactors is very extensive. Performance over a range of power and burnup levels for various operating conditions is quite satisfactory. However significant development programs are presently in progress to further extend our knowledge under increasingly severe operating conditions. In particular, upcoming data acquisition programs (1985-1988) will cover site and hot cell measurements on Gd poison rods, 4.5 % overenriched fuel rods over four operating cycles, 17 x 17 AFA fuel assemblies. For these products the same surveillance strategy as the one used for the standard assembly has been adopted, in order to continuously provide more data which can be used to upgrade design models and pave the way for the development of future products

  18. Characteristics of welded joints of nuclear reactor interest

    International Nuclear Information System (INIS)

    1978-01-01

    The main propose of this work, was the determination of the optical conditions for obtaining welded joints of stainless steel, the quality control of joints obtained by destructive and non-destructive essays, as well as, the first specific essays of fluence and fatigue of the base metals employed. All tests performed in the base metals are very important from the joint of view that the comparison between results obtained with base metals and welded joints allows a the evaluation of the efficiency of the welded joints. (author) [pt

  19. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  20. Evaluation of fuel performance for fresh and aged CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Bae, Jun Ho; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

  1. Reflector Performance Study in Ultra-long Cycle Fast Reactor

    International Nuclear Information System (INIS)

    Tak, Taewoo; Kong, Chidong; Choe, Jiwon; Lee, Deokjung

    2013-01-01

    There are reflector assemblies outside the fuel region, surrounding the fuel assemblies and axial reflector is located at the bottom of the core to control the neutron leakage fraction which is an important factor in fast reactor system. HT-9 was used as a reflector material as well as a structure material. In this study, alternative reflector materials were proposed and their reflection performance was tested and studied focused on its physics. ODS-MA957 and SiC were chosen from iron based alloy and ceramic respectively. The two materials were tested and compared with HT-9 in UCFR-1000 as a radial and an axial reflector and it was evaluated from the neutronics point of view with comparing the core life and the coolant void reactivity. The calculation and evaluation were performed by McCARD Monte Carlo code. The reflector materials for UCFR-1000 have been investigated in the aspect of neutronics. The reflection effect shows different performance corresponding to reflector material used. Also, the neutron energy spectrum is affected by changing materials which causes spectrum softening but it is not enough to influence the core life. With more reflector material candidates such as lead-based liquid metal, reflection performance and core parameter study will be investigated for next step

  2. Thermal characteristics analysis of microwaves reactor for pyrolysis of used cooking oil

    Science.gov (United States)

    Anis, Samsudin; Shahadati, Laily; Sumbodo, Wirawan; Wahyudi

    2017-03-01

    The research is objected to develop microwave reactor for pyrolysis of used cooking oil. The effect of microwave power as well as addition of char as absorber towards its thermal characteristic were investigated. Domestic microwave was modified and used to test the thermal characteristic of used cooking oil in the terms of temperature evolution, heating rate, and thermal efficiency. The samples were examined under various microwave power of 347W, 399W, 572W and 642W for 25 minutes of irradiation time. The char loading was tested in the level of 0, 50, and 100 g. Microwave reactor consists of microwave unit with a maximum power of 642W, a ceramic reactor, and a condenser equipped with temperature measurement system was successfully developed. It was found that microwave power and addition of absorber significantly influenced the thermal characteristic of microwave reactor. Under investigated condition, the optimum result was obtained at microwave power of 642W and 100 g of char. The condition was able to provide temperature of 480°C, heating rate of 18.2°C/min and thermal efficiency of 53% that is suitable to pyrolyze used cooking oil.

  3. Non-linearity consideration when analyzing reactor noise statistical characteristics. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kebadze, B V; Adamovski, L A

    1975-06-01

    Statistical characteristics of boiling water reactor noise in the vicinity of stability threshold are studied. The reactor is considered as a non-linear system affected by random perturbations. To solve a non-linear problem the principle of statistical linearization is used. It is shown that the halfwidth of resonance peak in neutron power noise spectrum density as well as the reciprocal of noise dispersion, which are used in predicting a stable operation theshold, are different from zero both within and beyond the stability boundary the determination of which was based on linear criteria.

  4. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  5. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

    1998-01-01

    A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

  6. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  7. Characteristics and economy of the European reactor of pressurized water (EPR)

    International Nuclear Information System (INIS)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C.

    2005-01-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  8. Characteristics of self-powered neutron detectors used in power reactors

    International Nuclear Information System (INIS)

    Todt, William H. Sr.

    1998-01-01

    Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)

  9. Experimental investigation of the neutron physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Thong, Ha Van [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The investigation of the neutron physics characteristics of the Dalat Reactor has obtained the results as follows: 1/ The effective fraction of delayed photoneutrons and the extraneous neutron source left after reactor shut down are measured. 2/ The lowest power levels of critical states of the reactor are determined. 3/The perturbation effect is investigated when a water column or a plexiglass rod is substituted for a fuel element. 4/ The relative axial and radial distributions of the thermal neutrons measured and the geometrical parameters of the core such as the inhomogeneous coefficients, the buckling, the effective height and radius, the extrapolated distances are obtained. 4/ The thermal neutron distributions are measured around the old graphite reflector. (author). 10 refs., 10 figs., 2 tabs.

  10. Calculation of static characteristics of linear step motors for control rod drives of nuclear reactors - an approximate approach

    International Nuclear Information System (INIS)

    Khan, S.H.; Ivanov, A.A.

    1993-01-01

    This paper describes an approximate method for calculating the static characteristics of linear step motors (LSM), being developed for control rod drives (CRD) in large nuclear reactors. The static characteristic of such an LSM which is given by the variation of electromagnetic force with armature displacement determines the motor performance in its standing and dynamic modes. The approximate method of calculation of these characteristics is based on the permeance analysis method applied to the phase magnetic circuit of LSM. This is a simple, fast and efficient analytical approach which gives satisfactory results for small stator currents and weak iron saturation, typical to the standing mode of operation of LSM. The method is validated by comparing theoretical results with experimental ones. (Author)

  11. An analytical method for the calculation of static characteristics of linear step motors for control rod drives in nuclear reactors

    International Nuclear Information System (INIS)

    Khan, S.H.; Ivanov, A.A.

    1995-01-01

    An analytical method for calculating static characteristics of linear dc step motors (LSM) is described. These multiphase passive-armature motors are now being developed for control rod drives (CRD) in large nuclear reactors. The static characteristics of such LSM is defined by the variation of electromagnetic force with armature displacement and it determines motor performance in its standing and dynamic modes of operation. The proposed analytical technique for calculating this characteristic is based on the permeance analysis method applied to phase magnetic circuits of LSM. Reluctances of various parts of phase magnetic circuit is calculated analytically by assuming probable flux paths and by taking into account complex nature of magnetic field distribution in it. For given armature positions stator and armature iron saturations are taken into account by an efficient iterative algorithm which gives fast convergence. The method is validated by comparing theoretical results with experimental ones which shows satisfactory agreement for small stator currents and weak iron saturation

  12. Assessment of beam tube performance for the maple research reactor

    International Nuclear Information System (INIS)

    Lee, A.G.

    1986-06-01

    The MAPLE research reactor is a versatile new research facility that can be adapted to meet the requirements of a variety of reactor applications. A particular group of reactor applications involves the use of beams of radiation extracted from the reactor core via tubes that penetrate through the biological shield and terminate in the reflector surrounding the fuelled core. An assessment is given of the neutron and gamma radiation fields entering beam tubes that are located radially or tangentially with respect to the core

  13. Design characteristics of metallic fuel rod on its in-LMR performance

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

  14. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  15. THESEE-3, Orgel Reactor Performance and Statistic Hot Channel Factors

    International Nuclear Information System (INIS)

    Chambaud, B.

    1974-01-01

    1 - Nature of physical problem solved: The code applies to a heavy-water moderated organic-cooled reactor channel. Different fuel cluster models can be used (circular or hexagonal patterns). The code gives coolant temperatures and velocities and cladding temperatures throughout the channel and also channel performances, such as power, outlet temperature, boiling and burn-out safety margins (see THESEE-1). In a further step, calculations are performed with statistical values obtained by random retrieval of geometrical in- put data and taking into account construction tolerances, vibrations, etc. The code evaluates the mean value and standard deviation for the more important thermal and hydraulic parameters. 2 - Method of solution: First step calculations are performed for nominal values of parameters by solving iteratively the non-linear system of equations which give the pressure drops in subchannels of the current zone (see THESEE-1). Then a Gaussian probability distribution of possible statistical values of the geometrical input data is assumed. A random number generation routine determines the statistical case. Calculations are performed in the same way as for the nominal case. In the case of several channels, statistical performances must be adjusted to equalize the normal pressure drop. A special subroutine (AVERAGE) then determines the mean value and standard deviation, and thus probability functions of the most significant thermal and hydraulic results. 3 - Restrictions on the complexity of the problem: Maximum 7 fuel clusters, each divided into 10 axial zones. Fuel bundle geometries are restricted to the following models - circular pattern 6/7, 18/19, 36/67 rods, with or without fillers. The fuel temperature distribution is not studied. The probability distribution of the statistical input is assumed to be a Gaussian function. The principle of random retrieval of statistical values is correct, but some additional correlations could be found from a more

  16. Nuclear power plant design characteristics. Structure of nuclear power plant design characteristics in the IAEA Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    2007-03-01

    One of the IAEA's priorities has been to maintain the Power Reactor Information System (PRIS) database as a viable and useful source of information on nuclear reactors worldwide. To satisfy the needs of PRIS users as much as possible, the PRIS database has included also a set of nuclear power plant (NPP) design characteristics. Accordingly, the PRIS Technical Meeting, organized in Vienna 4-7 October 2004, initiated a thorough revision of the design data area of the PRIS database to establish the actual status of the data and make improvements. The revision first concentrated on a detailed review of the design data completion and the composition of the design characteristics. Based on the results of the review, a modified set and structure of the unit design characteristics for the PRIS database has been developed. The main objective of the development has been to cover all significant plant systems adequately and provide an even more comprehensive overview of NPP unit designs stored in the PRIS database

  17. Prediction of Hydraulic Performance of a Scaled-Down Model of SMART Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Sun Guk; Park, Jin Seok; Yu, Je Yong; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-08-15

    An analysis was conducted to predict the hydraulic performance of a reactor coolant pump (RCP) of SMART at the off-design as well as design points. In order to reduce the analysis time efficiently, a single passage containing an impeller and a diffuser was considered as the computational domain. A stage scheme was used to perform a circumferential averaging of the flux on the impeller-diffuser interface. The pressure difference between the inlet and outlet of the pump was determined and was used to compute the head, efficiency, and break horse power (BHP) of a scaled-down model under conditions of steady-state incompressible flow. The predicted curves of the hydraulic performance of an RCP were similar to the typical characteristic curves of a conventional mixed-flow pump. The complex internal fluid flow of a pump, including the internal recirculation loss due to reverse flow, was observed at a low flow rate.

  18. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

  19. Characteristics of self-powered neutron detectors used in power reactors

    International Nuclear Information System (INIS)

    Todt, W.H.

    1997-01-01

    Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

  20. Core characteristics on a hybrid type fast reactor system combined with proton accelerator

    International Nuclear Information System (INIS)

    Kowata, Yasuki; Otsubo, Akira

    1997-06-01

    In our study on a hybrid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleus has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long term without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100 cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year. (author)

  1. Intelligent information database of the thermal-hydraulic characteristics for a future marine water reactor

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki

    2000-01-01

    At the Ship Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated type marine water reactor has been conducted. This current study aims at developing an intelligent information database program with the thermal-hydraulic characteristics of a future marine water reactor on the basis of the valuably experimental knowledge, which was obtained from the above-mentioned studies. In this paper, the experimental knowledge with the flow boiling of a once-through steam generator and the natural circulation of primary water under a ship rolling motion was converted into an intelligent information database program. The program was created as a Windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability for any helical-coil type once-through steam generator design, (2) reference and graphic display of the experimental data, (3) reference of the information such as analysis method and experimental apparatus. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized reactor with helical-coil type steam generator. (author)

  2. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  3. Destination Characteristics that Drive Hotel Performance

    DEFF Research Database (Denmark)

    Assaf, A. George; Josiassen, Alexander; Woo, Linda

    2017-01-01

    , government support, disposable income, and number of international arrivals within a tourism destination. Results indicate that the most important barriers to hotel performance are the competition among accommodation providers, tax rate and fuel price. We argue for the need for hotel providers to develop......The increased market saturation and competition in both domestic and international tourism destinations have renewed interest among hotel operators in identifying the key drivers of hotel performance. This paper presents a comprehensive analysis of the determinants of hotel performance...... strategies that take cognisance of the key drivers and barriers to enhancing hotel performance in an ever-changing global tourism sector....

  4. Operating performance of the prototype heavy water reactor Fugen

    International Nuclear Information System (INIS)

    1984-01-01

    Since the full scale operation was started in March, 1979, the ATR Fugen power station has been verifying the performance and reliability of the machinery and equipment, uranium-plutonium mixed oxide fuel and so on, and obtaining the technical prospect for putting ATRs in practical use by accumulating operation and maintenance techniques, through about five years of operation. In this report, the operational results of the Fugen power station are described. Fugen is a heavy water-moderated, boiling light water-cooled, pressure tube type reactor with 165 MWe output. As of the end of March, 1984, the total generated electric power was about 4.3 billion kWh, and the operation time was about 27,000 hours. The mean capacity ratio reached 58.8%. During the operation period, troubles including plant shutdown occurred eight times, but generally the performance and reliability of the machinery and equipment have been good. 580 fuels including 284 MOX fuels have been charged, but fuel breaking did not occur at all. The consumption of heavy water and the leak of tritium did not cause problem. The management of the core and fuel, the management of maintenance, the quality control of cooling water and heavy water, radiation control and the management of wastes are reported. (Kako, I.)

  5. Light water reactor pressure isolation valve performance testing

    International Nuclear Information System (INIS)

    Neely, H.H.; Jeanmougin, N.M.; Corugedo, J.J.

    1990-07-01

    The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (author)

  6. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  7. Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some

  8. Investigation on discharge characteristics of a coaxial dielectric barrier discharge reactor driven by AC and ns power sources

    Science.gov (United States)

    Qian, WANG; Feng, LIU; Chuanrun, MIAO; Bing, YAN; Zhi, FANG

    2018-03-01

    A coaxial dielectric barrier discharge (DBD) reactor with double layer dielectric barriers has been developed for exhaust gas treatment and excited either by AC power or nanosecond (ns) pulse to generate atmospheric pressure plasma. The comparative study on the discharge characteristics of the discharge uniformity, power deposition, energy efficiency, and operation temperature between AC and ns pulsed coaxial DBD is carried out in terms of optical and electrical characteristics and operation temperature for optimizing the coaxial DBD reactor performance. The voltages across the air gap and dielectric layer and the conduction and displacement currents are extracted from the applied voltages and measured currents of AC and ns pulsed coaxial DBDs for the calculation of the power depositions and energy efficiencies through an equivalent electrical model. The discharge uniformity and operating temperature of the coaxial DBD reactor are monitored and analyzed by optical images and infrared camera. A heat conduction model is used to calculate the temperature of the internal quartz tube. It is found that the ns pulsed coaxial DBD has a much higher instantaneous power deposition in plasma, a lower total power consumption, and a higher energy efficiency compared with that excited by AC power and is more homogeneous and stable. The temperature of the outside wall of the AC and ns pulse excited coaxial DBD reaches 158 °C and 64.3 °C after 900 s operation, respectively. The experimental results on the comparison of the discharge characteristics of coaxial DBDs excited by different powers are significant for understanding of the mechanism of DBDs, reducing energy loss, and optimizing the performance of coaxial DBD in industrial applications.

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

    International Nuclear Information System (INIS)

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

    1987-05-01

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

  10. Performance of nuclear fuel in the Krsko reactor; Spremljanje delovanja jedrskega goriva v reaktorju NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Jurcevic, M; Kurincic, B; Levstek, M F; Sambo, B; Vrcko, P [Nuklearna elektrana Krsko, Krsko (Yugoslavia)

    1987-07-01

    In this paper activities to follow performance of the nuclear fuel and operational status of the reactor of Nuclear Power Plant Krsko are presented. Short descriptions of the methods as well as nuclear and process instrumentation used for surveillance of the reactor performance are given. The purpose of the subject activities is to assure safe operation of the reactor in accordance with the Final safety Analysis Report of NPP Krsko. (author)

  11. International experience and status of fuel element performance and modelling for water reactors

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    Current knowledge concerning water reactor fuel performance and technology is reviewed (212 references). The emphasis is on aspects of in-reactor performance including behaviour in accidents. Computer models for predicting fuel behaviour during the ordinary running of the reactor and during accidents are described. These codes include COMETHE, HOTROD, SLEUTH-SEER and FRAPCON. Their agreement with experimental data is examined. (U.K.)

  12. Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Bang, In Cheol

    2017-01-01

    Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

  13. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  14. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  15. Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2015-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

  16. Haematological characteristics and performance of West African ...

    African Journals Online (AJOL)

    STORAGESEVER

    2009-02-18

    Feb 18, 2009 ... ISSN 1684–5315 © 2009 Academic Journals. Full Length ... performance of 36 young West African Dwarf (WAD) goats was investigated in order to simulate the .... cause antibody depression, alter white blood cell counts,.

  17. Sire influence on reproductive, performance characteristics and ...

    African Journals Online (AJOL)

    , Panda White x Cinnamon Brown (PWxCB) and Silver Brown x Cinnamon Brown (SBxCB). The experiment was a randomized complete block design. Parameters measured include: fertility and hatchability traits, growth performance traits and ...

  18. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    International Nuclear Information System (INIS)

    Tukiran Surbakti

    2009-01-01

    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

  19. PLUS7TM In-Reactor Operating Performance and Economics

    International Nuclear Information System (INIS)

    Kim, Kyutae; Jang, Youngki; Choi, Joonhyung; Lee, Jinseok; Kim, Yoonho; Suh, Jungmin

    2006-01-01

    KNFC has developed an advanced fuel, PLUS7 TM , for the Korean Standard Nuclear Power Plants(KSNPs) through the joint development program with Westinghouse. With the help of various out-of-pile tests, it is found that the PLUS7 TM shows much better performance than the current fuel, GUARDIAN TM from the safety and economy points of view. Now four Lead Test Assembles(LTAs) of the PLUS7 TM are being irradiated for the 3 rd cycle after the successful completion of the 1 st and 2 nd irradiation cycles. During the 1 st and 2 nd irradiation cycles, no fuel failure was observed at LTAs and their nuclear-related parameters matched their design values well. During the overhaul period, on the other hand, pool side examinations were performed for four LTAs to generate key in-reactor fuel performance data such as fuel rod and assembly growths, fuel rod-to-top nozzle gap, fuel assembly bow and twist, fuel rod bow, spacer grid width, fuel rod diameter and fuel rod oxide layer thickness. It is found that all measured values are bounded by upper and lower predicted ones. The detailed economic analyses have shown that significant fuel cycle cost can be reduced by more than one million dollars per cycle of one KSNP with the introduction of the PLUS7 TM assembly. Furthermore, more than one hundred million dollars with power up-rating of 5% can be saved annually for currently operating eight KSNPs, which is easily and safety achievable with the PLUS7 TM assembly

  20. Performance Assessment of Turbulence Models for the Prediction of the Reactor Internal Flow in the Scale-down APR+

    International Nuclear Information System (INIS)

    Lee, Gonghee; Bang, Youngseok; Woo, Swengwoong; Kim, Dohyeong; Kang, Minku

    2013-01-01

    The types of errors in CFD simulation can be divided into the two main categories: numerical errors and model errors. Turbulence model is one of the important sources for model errors. In this study, in order to assess the prediction performance of Reynolds-averaged Navier-Stokes (RANS)-based two equations turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. In this study, in order to assess the prediction performance of turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. Both standard k-ε model and SST model predicted the similar flow pattern inside reactor. Therefore it was concluded that the prediction performance of both turbulence models was nearly same. Complex thermal-hydraulic characteristics exist inside reactor because the reactor internals consist of fuel assembly, control rod assembly, and the internal structures. Either flow distribution test for the scale-down reactor model or computational fluid dynamics (CFD) simulation have been conducted to understand these complex thermal-hydraulic features inside reactor

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  2. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  3. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wood, J C; Foo, M T; Berthiaume, L C; Herbert, L N; Schaefer, J D; Hawley, D [Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, ON KOJ 1JO (Canada)

    1985-07-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U{sub 3}Si in aluminum, to complement the dispersions of U{sub 3}Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U{sub 3}Si have been manufactured. (author)

  4. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.; Hawley, D.

    1985-01-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U 3 Si in aluminum, to complement the dispersions of U 3 Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U 3 Si have been manufactured. (author)

  5. Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

    International Nuclear Information System (INIS)

    Hejzlar, Pavel; Davis, Cliff B.

    2004-01-01

    A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners

  6. Data characteristics that determine classifier performance

    CSIR Research Space (South Africa)

    Van der Walt, Christiaan M

    2006-11-01

    Full Text Available available at [11]. The kNN uses a LinearNN nearest neighbour search algorithm with an Euclidean distance metric [8]. The optimal k value is determined by performing 10-fold cross-validation. An optimal k value between 1 and 10 is used for Experiments 1... classifiers. 10-fold cross-validation is used to evaluate and compare the performance of the classifiers on the different data sets. 3.1. Artificial data generation Multivariate Gaussian distributions are used to generate artificial data sets. We use d...

  7. A study on the change in dynamic characteristics of reactor internals

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Jung, Seung Ho; Park, Jin Ho; Park, Jin Suk; Jeong, Keong Hoon

    1993-12-01

    An experimental and analytical studies were performed to establish the relationship between in-air dynamic characteristics of reactor internals and in-water ones by using a scaled-down model which consists of CSB (flexible) and RPV (rigid) models. The experimental results show that the natural frequencies of the CSB model in water are remarkably lowered than those in air, and the normalized natural frequency (in-water frequency/in-air frequency) of the CSB model exists between in-phase mode and out-of-phase mode values of the flexible-to-flexible co-axial cylindrical structure which has the same dimensions as the scaled-down model. The normalized frequency increases to asymptotically reach the in-phase mode frequency of the flexible co-axial cylindrical structure as circumferential mode number (n) increases, while decreases to come near of the out-of-phase mode value as the circumferential mode number decreases. The support condition change for the CSB model in waster made the frequency of 1st axial mode (pure beam mode) shifted into lower value without any effect on the other modes. The changing trend of the natural frequencies and the mode shapes resulted from the finite element analysis by using the ANSYS code shows good agreement with the experimental results both in air and in water cases. In addition, a theoretical study was also performed by simplifying the scaled-down model as flexible-to-rigid co-axial cylindrical structure. It is found that the in-air natural frequencies shows good agreement with both the experimental and finite element analysis results, while the in-water frequencies reveals rather discrepancies. (Author)

  8. performance characteristics of a cam turning attachment

    African Journals Online (AJOL)

    Dr Obe

    ABSTRACT. A modification of a cylindrical turning unit has been done to give a non- cylindrical turning attachment for production of irregular shapes, like cams on the lathe machine. To assess the performance of the attachment, cutting forces have been measured using a 'Sigma' Cutting Tool. Dynamometer. Furthermore ...

  9. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  10. Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van

    2006-01-01

    gradients within the fuel assemblies would be too high, and fuel economy is poor. Two improved fuel concepts are proposed: (1) a redesign of the classic TRISO coated particle fuel, and (2) an innovative hollow sphere design. Both fuel elements are used in a core design based on direct cooling of the coated particle fuel. To increase the neutronic margins and obtain adequate self-breeding capabilities, the proposed reactor has 2400 MWth power output and a power density of 50 MW/m 3 . With both types of fuel, it is possible to obtain a closed fuel cycle. Long irradiation intervals (several years) are possible with a low burnup reactivity swing, which reduces the required over-reactivity of the fresh core and reduces control rod requirements during operation. In the closed fuel cycle it is important to be able to predict whether a certain initial fuel composition will in fact yield a new fuel, after irradiation, cool down and reprocessing, with which the reactor can be restarted. A theoretical framework is presented in this thesis which allows calculation of the ‘Breeding Gain’ (BG) of the reactor. The BG quantifies the performance of the fuel for batch i + 1 as a function of the composition of the initial fuel of batch i. If this BG can be made equal to zero, both fuel compositions give the same nuclear performance. To be able to calculate the fuel performance, the reactivity weight, i.e. the contribution of each isotope to the overall reactivity of the reactor, needs to be estimated. It is proposed in this thesis to calculate these reactivity weights using a first-order eigenvalue perturbation calculation. It is shown that this approach yields an expression which reduces to a well-established formula for reactivity weights. All steps in the fuel cycle, i.e. irradiation, cool down and reprocessing, have to be taken into account to calculate the Breeding Gain for the closed fuel cycle. First order nuclide perturbation theory provides an efficient method to calculate the

  11. Operation and Performance of the Supercritical Fluids Reactor (SFR)

    National Research Council Canada - National Science Library

    Hanush, R

    1996-01-01

    The Supercritical Fluids Reactor (SFR) at Sandia National Laboratories, CA has been developed to examine and solve engineering, process, and fundamental chemistry issues regarding the development of supercritical water oxidation (SCWO...

  12. Performance characteristics of a shower cooling tower

    International Nuclear Information System (INIS)

    Qi Xiaoni; Liu Zhenyan; Li Dandan

    2007-01-01

    This study was prompted by the need to design towers for applications in which, due to salt deposition on the packing and subsequent blockage, the use of tower packing is not practical. In contrast to conventional cooling towers, the cooling tower analyzed in this study is void of fill. By means of efficient atomization nozzles, a shower cooling tower (SCT) is possible to be applied in industry, which, in terms of water cooling, energy saving and equipment investing, is better than conventional packed cooling towers. However, no systematic thermodynamic numerical method could be found in the literature up to now. Based on the kinetic model and mass and heat transfer model, this paper has developed a one dimensional model for studying the motional process and evaporative cooling process occurring at the water droplet level in the SCT. The finite difference approach is used for three motional processes to obtain relative parameters in each different stage, and the possibility of the droplets being entrained outside the tower is fully analyzed. The accuracy of this model is checked by practical operational results from a full scale prototype in real conditions, and some exclusive factors that affect the cooling characteristics for the SCT are analyzed in detail. This study provides the theoretical foundation for practical application of the SCT in industry

  13. Some useful characteristics of performance models

    International Nuclear Information System (INIS)

    Worledge, D.H.

    1985-01-01

    This paper examines the demands placed upon models of human cognitive decision processes in application to Probabilistic Risk Assessment. Successful models, for this purpose, should, 1) be based on proven or plausible psychological knowledge, e.g., Rasmussen's mental schematic, 2) incorporate opportunities for slips, 3) take account of the recursive nature, in time, of corrections to mistaken actions, and 4) depend on the crew's predominant mental states that accompany such recursions. The latter is equivalent to an explicit coupling between input and output of Rasmussen's mental schematic. A family of such models is proposed with observable rate processes mediating the (conscious) mental states involved. It is expected that the cumulative probability distributions corresponding to the individual rate processes can be identified with probability-time correlations of the HCR Human Cognitive Reliability type discussed elsewhere in this session. The functional forms of the conditional rates are intuitively shown to have simple characteristics that lead to a strongly recursive stochastic process with significant predictive capability. Models of the type proposed have few parts and form a representation that is intentionally far short of a fully transparent exposition of the mental process in order to avoid making impossible demands on data

  14. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    International Nuclear Information System (INIS)

    Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

    2012-01-01

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

  15. Comparison of some characteristics of aerobic granules and sludge flocs from sequencing batch reactors.

    Science.gov (United States)

    Li, J; Garny, K; Neu, T; He, M; Lindenblatt, C; Horn, H

    2007-01-01

    Physical, chemical and biological characteristics were investigated for aerobic granules and sludge flocs from three laboratory-scale sequencing batch reactors (SBRs). One reactor was operated as normal SBR (N-SBR) and two reactors were operated as granular SBRs (G-SBR1 and G-SBR2). G-SBR1 was inoculated with activated sludge and G-SBR2 with granules from the municipal wastewater plant in Garching (Germany). The following major parameters and functions were measured and compared between the three reactors: morphology, settling velocity, specific gravity (SG), sludge volume index (SVI), specific oxygen uptake rate (SOUR), distribution of the volume fraction of extracellular polymeric substances (EPS) and bacteria, organic carbon and nitrogen removal. Compared with sludge flocs, granular sludge had excellent settling properties, good solid-liquid separation, high biomass concentration, simultaneous nitrification and denitrification. Aerobic granular sludge does not have a higher microbial activity and there are some problems including higher effluent suspended solids, lower ratio of VSS/SS and no nitrification at the beginning of cultivation. Measurement with CLSM and additional image analysis showed that EPS glycoconjugates build one main fraction inside the granules. The aerobic granules from G-SBR1 prove to be heavier, smaller and have a higher microbial activity compared with G-SBR2. Furthermore, the granules were more compact, with lower SVI and less filamentous bacteria.

  16. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  17. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  18. Startup and operating characteristics of an external air-lift reflux partial nitritation-ANAMMOX integrative reactor.

    Science.gov (United States)

    Li, Xiang; Huang, Yong; Yuan, Yi; Bi, Zhen; Liu, Xin

    2017-08-01

    The differences in the physiological characteristics between AOB and ANAMMOX bacteria lead to suboptimal performance when used in a single reactor. In this study, aerobic and anaerobic zones with different survival environments were constructed in a single reactor to realize partitioned culture of AOB and ANAMMOX bacteria. An external air-lift reflux system was formed which used the exhaust from the aeration zone as power to return the effluent to the aeration zone. The reflux system effectively alleviated the large pH fluctuations and promoted NO 2 - -N to rapidly use by ANAMMOX bacteria, effectively inhibiting the activity of NOB. After 95d of running, the nitrogen removal rate increased from the initial 0.21kg/(m 3 ·d) to 3.1kg/(m 3 ·d). FISH analyses further demonstrated that AOB and ANAMMOX bacteria acquired efficient enrichment in the corresponding zone. Thus, this type of integrative reactor may create the environments needed for the partial nitritation-ANAMMOX processing. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Performance and carcass characteristics of Yankasa ram fed with ...

    African Journals Online (AJOL)

    Remember me ... and 50% maize and wheat offal mixture, were better when compared to the control (B0) and other test diet in terms of performance and carcass characteristics. ... Key words: Performance, carcass, biscuit waste, Yankasa ram.

  20. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  1. Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

    International Nuclear Information System (INIS)

    Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

    2013-01-01

    The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

  2. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  3. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  4. Performance characteristics for advanced control systems

    International Nuclear Information System (INIS)

    Kisner, R.A.

    1989-01-01

    A growing collection of control techniques is becoming available to the design engineer. This make selection of the most appropriate technique for a given application a difficult task. A systematic approach to evaluating alternative control schemes is needed. The approach discussed in this paper expands the traditional concepts of quantitative performance analysis to include other relevant factors such as robustness of the technique, resource requirements, and effects on operators and other personnel. This collection of factors, termed measures of utility, may be used as qualitative and quantitative means of evaluating and comparing properties of alternative control system designs. This paper, although not an in-depth study, serves to outline several measures of utility and suggests a general structure for control system development. This method of comparing the usefulness of alternative control system will prove valuable to the ORNL Advanced Controls Program (ACTO) for optimizing compatibility with actual systems and equipment

  5. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  6. Comparison of fuel cycles characteristics for nuclear energy systems based on WWER-TOI and BN-1200 reactors

    International Nuclear Information System (INIS)

    Kagramanyan, V.S.; Kalashnikov, A.G.; Kapranova, Eh.N.; Puzakov, A.Yu.

    2014-01-01

    Authors determine the characteristics of the fuel cycle (FC) based on stationary nuclear power system based on WWER-TOI and BN-1200 reactors with fuel of different composition. Characteristics of reactor systems with partial or complete spent nuclear fuel reprocessing and recycling of plutonium are compared to those of the reference system consisting only of WWER-TOI with uranium oxide fuel, operating in an open FC [ru

  7. Biomass characteristics in three sequencing batch reactors treating a wastewater containing synthetic organic chemicals

    DEFF Research Database (Denmark)

    Hu, Z.Q.; Ferraina, R.A.; Ericson, J.F.

    2005-01-01

    in all reactors. In contrast, effluent 3-nitrobenzoate was recorded when its influent concentration was increased to 5 mg L-1 and dropped only to below 1 mg L-1 after 300 days of operation. The competent (active) biomass fractions for these compounds were between 0.04% and 5.52% of the total biomass...... characteristics in the aerobic SBR and SBBR. While all reactors had very good COD removal (> 90%) and displayed nitrification, substantial nitrogen removal (74%) was only achieved in the anoxic/aerobic SBR. During the entire operational period, benzoate, theophylline and 4-chlorophenol were completely removed...... inferred from substrate-specific microbial enumerations. The measured competent biomass fractions for 4-chlorophenol and 3-nitrobenzoate degradation were significantly lower than the influent COD fractions of these compounds. Correspondent to the highest competent biomass fraction for benzoate degradation...

  8. Experimental and calculational works on characteristics of the Dalat Nuclear Research Reactor. Second edition

    International Nuclear Information System (INIS)

    Pham Ngoc Khoi; Nguyen Kim Dung

    2016-03-01

    Recognizing the significant value and necessity of publishing the scientific document of experimental and calculational works on the Dalat Nuclear Research Reactor (DNRR) physics and engineering for research, operation, training activities as well as for international scientific exchange, Vietnam Atomic Energy Agency (VAEA) and Vietnam Atomic Energy Institute have completed editing to publish the “Experimental and Calculational Works on Characteristics of THE DALAT NUCLEAR RESEARCH REACTOR” which consists of 26 typical papers representing the most important experimental and calculational results of the DNRR physics and engineering obtained during 30 years of operation and exploitation with the contribution of Vietnamese and former USSR’s experts, especially scientists and engineers working at the Reactor Center of the NRI

  9. Neutron characteristics of the Super-Phenix 1 reactor at Creys-Malville

    International Nuclear Information System (INIS)

    Giacometti, C.; Bouget, Y.H.; Hammer, P.; Lyon, F.; Salvatores, M.; Sicard, B.; Pipaud, J.Y.

    1980-01-01

    The paper describes the method used to determine the critical enrichments for the first loading of the Super-Phenix reactor and the correction factors (together with their uncertainties) applied to the data calculated from the CARNAVAL IV code. These enrichments must be chosen so as to conform to the planned operating conditions of the reactor: nominal power of the pressure vessels, lifetime of the in-pile assemblies. Allowance for uncertainties of neutronic origin and those associated with the fabrication of the fuel pins calls for an over-enrichment of the first loading by approximately 4 per cent. An analysis is made of the effects of this over-enrichment on the core characteristics, which have to remain compatible with the established limits. (author)

  10. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  11. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  12. Basic Characteristics of Human Erroneous Actions during Test and Maintenance Activities Leading to Unplanned Reactor Trips

    International Nuclear Information System (INIS)

    Kim, Jae Whan; Park, Jin Kyun

    2010-01-01

    Test and maintenance (T and M) activities of nuclear power plants are essential for sustaining the safety of a power plant and maintaining the reliability of plant systems and components. However, the potential of human errors during T and M activities has also the potential to induce unplanned reactor trips or power derate or making safety-related systems unavailable. According to the major incident/accident reports of nuclear power plants in Korea, contribution of human errors takes up about 20% of the total events. The previous study presents that most of human-related unplanned reactor trip events during normal power operation are associated with T and M activities (63%), which are comprised of plant maintenance activities such as a 'periodic preventive maintenance (PPM)', a 'planned maintenance (PM)' and a 'corrective maintenance (CM)'. This means that T and M activities should be a major subject for reducing the frequency of human-related unplanned reactor trips. This paper aims to introduce basic characteristics of human erroneous actions involved in the test and maintenance-induced unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants. The basic characteristics are described by dividing human erroneous actions into planning-based errors and execution-based errors. For the events associated with planning failures, they are, firstly, classified according to existence of the work procedure and then described for what aspects of the procedure or work plan have deficiency or problem. On the other hand, for the events associated with execution failures, they are described from the aspect of external error modes

  13. MODEL OF EMERGENCY DEPARTMENT NURSE PERFORMANCE IMPROVEMENT BASED ON ASSOCIATION OF INDIVIDUAL CHARACTERISTIC, ORGANIZATION CHARACTERISTIC AND JOB CHARACTERISTIC

    Directory of Open Access Journals (Sweden)

    Maria Margaretha Bogar

    2017-04-01

    Full Text Available Introduction: Nursing care is integral part of health care and having important role in management of patient with emergency condition. The purpose of this research was to develop nurse performance improvement model based on individual, organization and job characteristics association in Emergency Department of RSUD dr TC Hillers Maumere. Method: This was an explanative survey by cross sectional approach held on July -August 2012. Respondents in this study were 22 nurses and 44 patients were obtained by purposive sampling technique. Data were analyzed by partial least square test and signi fi cant t value > 1.64 (alpha 10%. Result: Results showed that individual characteristic had effect on nurse performance (t = 7.59, organization characteristic had effect on nurse performance (t = 2.03 and job characteristic didn’t have effect on nurse performance (t = 0.88. Nurse performance had effect on patient satisfaction (t = 6.54 but nurse satisfaction didn’t have effect on nurse performance (t = 1.31, and nurse satisfaction didn’t have effect either on patient satisfaction (t = 0.94. Discussion: This research concluded that individual characteristics which in fl uence nurse performance in nursing care were ability and skill, experience, age, sex, attitude and motivation. Organization characteristic that influence nurse performance was reward while job characteristic that include job design and feedback didn’t influence nurse performance in nursing care. Nurse performance influenced patient satisfaction but nurse satisfaction didn’t influence patient satisfaction and nurse performance.

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

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

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

  15. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

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

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

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

  17. Performance Testing of Hydrodesulfurization Catalysts Using a Single-Pellet-String Reactor

    NARCIS (Netherlands)

    Moonen, Roel; Ras, Erik Jan; Harvey, Clare; Alles, Jeroen; Moulijn, J.A.

    2017-01-01

    Small-scale parallel trickle-bed reactors were used to evaluate the performance of a commercial hydrodesulfurization catalyst under industrially relevant conditions. Catalyst extrudates were loaded as a single string in reactor tubes. It is demonstrated that product sulfur levels and densities

  18. Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

    International Nuclear Information System (INIS)

    Adlys, G.; Adliene, D.

    2002-01-01

    Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

  19. Passive devices of a reactor stop: classification of the characteristics and estimation of perfection degree

    International Nuclear Information System (INIS)

    Portyanoj, A.G.; Serdun', E.N.; Sorokin, A.P.; Egorov, V.S.; Shkarovskij, D.A.

    1998-01-01

    The perspective direction in NPP safety improvement connected with development of passive devices for nuclear reactor emergency shutdown (PDRS) is discussed. More than hundred devices which can fulfil the PDRS functions are suggested nowadays. The analysis of PDRS designing status as applicable for the fast reactors in the main which are based on the physical effect used in an element sensitive to temperature is made. The complex consisting of nine general characteristics including passive character, thresholdness, forces generation, inertia, multichannel design, stability towards operational factors, safety at failures, simplicity and visualisation, development conditions, is suggested for estimation of the quality of PDRS of different types. Basing on expert assessments realized using the complex of general characteristics it is shown that the types of PDRS may be separated into following three groups: linear expansion of solid bodies and thermoelectric ones (K ≅ 0.45); magnet ones with shape memory effect, liquid volume expansion (K ≅ 0.6); fusing ones (K ≅ 0.7). The conclusion is made that PDRS on the basis of fusing devices of the sulphon type with liofobic capillary-porous working body most completely satisfy the complex of general characteristics considered

  20. Effects of UO2 fuel microstructure and density on fuel in-reactor performance

    International Nuclear Information System (INIS)

    Hansson, L.

    1988-02-01

    The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

  1. Analysis of Topaz-II reactor performance using MCNP and TFEHX

    International Nuclear Information System (INIS)

    Lee, H.H.; Klein, A.C.

    1993-01-01

    Data reported by Russian scientist and engineers for the TOPAZ-II Space Nuclear Power is compared with analytical results calculated using the Monte Carlo Neutron and Photon (MCNP) and TFEHX computer codes. The results of these comparisons show good agreement with the TOPAZ-II neutronics, thermionic and thermal hydraulics performance. A detailed description of the TOPAZ-II reactor and of the TFE should enhance the performance of the both codes in modeling the reactor and TFE performances

  2. Thermal performance of an insulating structure for a reactor vessel

    International Nuclear Information System (INIS)

    Aranovitch, E.; Crutzen, S.; LeDet, M.; Denis, R.

    This report describes the installations used to test the HTGR reactor vessel insulating structure called ''Casali'' and details the experimental results in 3 groups: general experiments, systematic study, and technological experiments. The results obtained make it possible to satisfactorily predict the behavior of the structure in a practical application

  3. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    Perry, R.T.; Gorenflo, H.; Daenner, W.

    1976-01-01

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU) [de

  4. Experimental study of flow field characteristics on bed configurations in the pebble bed reactor

    International Nuclear Information System (INIS)

    Jia, Xinlong; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jia, Haijun; Jiang, Shengyao

    2017-01-01

    Highlights: • PTV study of flow fields of pebble bed reactor with different configurations are carried out. • Some criteria are proposed to quantify vertical velocity field and flow uniformity. • The effect of different pebble bed configurations is also compared by the proposed criteria. • The displacement thickness is used analogically to analyze flow field characteristics. • The effect of mass flow variation in the stagnated region of the funnel flow is measured. - Abstract: The flow field characteristics are of fundamental importance in the design work of the pebble bed high temperature gas cooled reactor (HTGR). The different effects of bed configurations on the flow characteristics of pebble bed are studied through the PTV (Particle Tracking Velocimetry) experiment. Some criteria, e.g. flow uniformity (σ) and mass flow level (α), are proposed to estimate vertical velocity field and compare the bed configurations. The distribution of the Δθ (angle difference between the individual particle velocity and the velocity vector sum of all particles) is also used to estimate the resultant motion consistency level. Moreover, for each bed configuration, the thickness of displacement is analyzed to measure the effect of the funnel flow zone based on the boundary layer theory. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity and other characteristics; and the sequence of levels of each estimation criterion is obtained for all bed configurations. In addition, a good design of the pebble bed configuration is suggested and these estimation criteria can be also applied and adopted in testing other geometry designs of pebble bed.

  5. Power probability density function control and performance assessment of a nuclear research reactor

    International Nuclear Information System (INIS)

    Abharian, Amir Esmaeili; Fadaei, Amir Hosein

    2014-01-01

    Highlights: • In this paper, the performance assessment of static PDF control system is discussed. • The reactor PDF model is set up based on the B-spline functions. • Acquaints of Nu, and Th-h. equations solve concurrently by reformed Hansen’s method. • A principle of performance assessment is put forward for the PDF of the NR control. - Abstract: One of the main issues in controlling a system is to keep track of the conditions of the system function. The performance condition of the system should be inspected continuously, to keep the system in reliable working condition. In this study, the nuclear reactor is considered as a complicated system and a principle of performance assessment is used for analyzing the performance of the power probability density function (PDF) of the nuclear research reactor control. First, the model of the power PDF is set up, then the controller is designed to make the power PDF for tracing the given shape, that make the reactor to be a closed-loop system. The operating data of the closed-loop reactor are used to assess the control performance with the performance assessment criteria. The modeling, controller design and the performance assessment of the power PDF are all applied to the control of Tehran Research Reactor (TRR) power in a nuclear process. In this paper, the performance assessment of the static PDF control system is discussed, the efficacy and efficiency of the proposed method are investigated, and finally its reliability is proven

  6. Shock resistance characteristic of a spiral symmetry stream anaerobic bio-reactor.

    Science.gov (United States)

    Chen, Xiaoguang; Dai, Ruobin; Xiang, Xinyi; Li, Gang; Xu, Zhengqi; Hu, Tao; Abdelgadir, Awad

    2016-01-01

    The shock resistance characteristic (SRC) of an anaerobic bioreactor characterizes the ability of the anaerobic community in the reactor to withstand violent change in the living environment. In comparison with an upflow anaerobic sludge blanket reactor (UASBR), the SRC of a spiral symmetry stream anaerobic bio-reactor (SSSAB) was systematically investigated in terms of removal efficiency, adsorption property, settling ability, flocculability and fluctuations in these parameters. A quantitative assessment method for SRC was also developed. The results indicated that the SSSAB showed better SRC than the UASBR. The average value (m value) of chemical oxygen demand removal rates of the SSSAB was 86.0%. The contact angle of granules in the SSSAB present gradient distribution, that is the m value of contact angle increasing from bottom (84.5°) to top (93.9°). The m value of the density at the upper and lower sections of the SSSAB were 1.0611 g·cm(-3) and 1.0423 g·cm(-3), respectively. The surface mean diameter of granules in the SSSAB increased from 1.164 to 1.292 mm during operation. The absolute m value of zeta potential of granular sludge at the upper and lower sections of the SSSAB were 40.4 mV and 44.9 mV, respectively. The weighted mean coefficient variance (C̅V̅) value indicated SSSAB was more stable than the UASBR.

  7. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Snelgrove, J.L.; Cornella, R.J.

    1983-01-01

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  8. Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

    International Nuclear Information System (INIS)

    Liu Kukui; Li Manchang; Tang Chuanbao

    1997-01-01

    In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

  9. Job Characteristics, Work Involvement, and Job Performance of Public Servants

    Science.gov (United States)

    Johari, Johanim; Yahya, Khulida Kirana

    2016-01-01

    Purpose: The primary purpose of this study is to assess the predicting role of job characteristics on job performance. Dimensions in the job characteristics construct are skill variety, task identity, task significance, autonomy and feedback. Further, work involvement is tested as a mediator in the hypothesized link. Design/methodology/approach: A…

  10. Analysis of characteristic performance curves in radiodiagnosis by an observer

    International Nuclear Information System (INIS)

    Kossovoj, A.L.

    1988-01-01

    Methods and ways of construction of performance characteristic curves (PX-curves) in roentgenology, their qualitative and quantitative estimation are described. Estimation of PX curves application for analysis of scintigraphic and sonographic images is presented

  11. Evaluation of the productive performance characteristics of red ...

    African Journals Online (AJOL)

    Evaluation of the productive performance characteristics of red tilapia ( Oreochromis sp.) injected with shark DNA into skeletal muscles and maintained diets containing different levels of probiotic and amino yeast.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

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

  14. Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel

    International Nuclear Information System (INIS)

    Unesaki, H.; Isaka, S.; Nakagome, Y.

    2006-01-01

    Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k ∞ and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O 2 fuels as well as UO 2 fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k ∞ of (U, Th)O 2 fuels was found to be significantly large compared to that of UO 2 fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)

  15. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  16. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

    2016-05-15

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  17. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    International Nuclear Information System (INIS)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

    2016-01-01

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  18. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  19. Performance of the Cascade inertial-confinement-fusion conceptual reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1984-01-01

    A 4.5-m-radius rotating fusion reactor made of silicon carbide and containing a moving 1-m-thick lithium-ceramic granular blanket can produce 3000 MW/sub t/. The blanket operates at high temperature (>1200 K) leading to gross plant efficiencies of up to 60% using a combined helium-gas turbine (Brayton cycle) with a vapor bottoming cycle

  20. Minor actinides transmutation performance in a fast reactor

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2016-01-01

    Highlights: • A method for calculating MA transmutation for individual nuclides has been proposed by introducing two formulas of the MA transmutation. One corresponds to the difference of MA amounts, and the other corresponds to the sum of the fission amounts and the plutonium production amounts. • Using the method the MA transmutation was calculated for Np-237 and Am-241 in a fast reactor. The burnup period was changed from 1 year to 12 year. • For the 1 year burnup a large amount of Am-242m, Cm-242 are produced from Am-241. The total MA transmutation amount increases with burnup time, but its gradient with respect to burnup time decreases after 9 years, and the transmutation amount by overall fission increases almost linearly with burnup time. • However, after the 6 year burnup the fission contribution became large because of the large production of Pu isotopes from the original Am-241. • In addition to the homogeneous loading of the MA nuclides into the cores, a heterogeneous loading of Am-241 to the blanket region was considered. - Abstract: Results obtained in the project named “Study on Minor Actinides Transmutation using Monju Data”, which has been sponsored by the Ministry of Education, Culture, Sports, Science and Technology in Japan (MEXT) are described. In order to physically understand transmutation of individual MA nuclides in fast reactors, a new method was developed in which the MAs transmutation is interpreted by two formulas. One corresponds to the difference of individual MA nuclides amounts before and after a burnup period, and the other is the sum of amount of fission of a relevant MA nuclide and the net plutonium production from the MA nuclide during a burnup period. The method has been applied to two fast reactors with MA fuels loaded in cores homogeneously and in a blanket region heterogeneously. Numerical results of MA transmutation for the two reactors are shown.

  1. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur

  2. Development of high performance core for large fast breeder reactors

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Watari, Yoshio.

    1982-01-01

    Subsequently to the fast breeder prototype reactor ''Monju'', the construction of a demonstration reactor with 1000 MWe output is planned. This research aims at the establishment of the concept of a large core with excellent fuel breeding property and safety for a demonstration and commercial reactors. For the purpose, the optimum specification of fuel design as a large core was clarified, and the new construction of a core was examined, in which a disk-shaped blanket with thin peripheral edge is introduced at the center of a core. As the result, such prospect was obtained that the time for fuel doubling would be 1/2, and the energy generated in a core collapse accident would be about 1/5 as compared with a large core using the same fuel as ''Monju''. Generally, as a core is enlarged, the rate of breeding lowers. If a worst core collapse accident occurs, the scale of accident will be very large in the case of a ''Monju'' type large core. In an unhomogeneous core, an internal blanket is provided in the core for the purpose of improving the breeding property and safety. Hitachi Ltd. developed the concept of a large core unhomogeneous in axial direction and proposed it. The research on the fuel design for a large core, an unhomogeneous core and its core collapse accident is reported. (Kako, I.)

  3. MOX fuel fabrication, in reactor performance and improvement

    International Nuclear Information System (INIS)

    Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

    1998-01-01

    In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

  4. The Influence of Top Management Team Characteristics on BPD Performance

    Directory of Open Access Journals (Sweden)

    Joy Elly Tulung

    2015-12-01

    Full Text Available Based on ”upper echelons theory”, this paper investigates the relation between top management team composition and BPD performance. For top management team characteristics, we employ age, level of education, background of education, gender, and functional background, while for measured the BPD performance we employ return on asset (ROA, return on equity (ROE, capital adequacy ratio (CAR, net interest margin (NIM, loan to deposit ratio (LDR, non-performing loan (NPL and operation expenses to operation income (BOPO. The results show that all characteristics have positive significant influences on BPD performance.

  5. Study of performances, stability and microbial characterization of a Sequencing Batch Biofilter Granular Reactor working at low recirculation flow.

    Science.gov (United States)

    De Sanctis, Marco; Beccari, Mario; Di Iaconi, Claudio; Majone, Mauro; Rossetti, Simona; Tandoi, Valter

    2013-02-01

    The Sequencing Batch Biofilter Granular Reactor (SBBGR) is a promising wastewater treatment technology characterized by high biomass concentration in the system, good depuration performance and low sludge production. Its main drawback is the high energy consumption required for wastewater recirculation through the reactor bed to ensure both shear stress and oxygen supply. Therefore, the effect of low recirculation flow on the long-term (38 months) performance of a laboratory scale SBBGR was studied. Both the microbial components of the granules, and their main metabolic activities were evaluated (heterotrophic oxidation, nitrification, denitrification, fermentation, sulphate reduction and methanogenesis). The results indicate that despite reduced recirculation, the SBBGR system maintained many of its advantageous characteristics. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. Elaboration by tape-casting and co-sintering of multilayer catalytic membrane reactor- performances

    International Nuclear Information System (INIS)

    Julian, A.

    2008-12-01

    This research deals with the increasing interest of the conversion of natural gas into liquid fuels (diesel, kerosene) using the Gas To Liquid (GTL) process. Within this context, Catalytic Membrane-based Reactors (CMR) would allow an improvement of the process efficiency and a reduction of investment and production costs with respect to the present technologies. They allow performing the separation of oxygen from air, and the conversion of natural gas into synthesis gas within a single step. After having highlighted the economical and technological advantages of using a ceramic membrane for the production of syngas (H 2 + CO 2 ), the author describes the protocols of synthesis of powders selected for the dense membrane and the porous support, and their physical characteristics. The obtained powders are then adapted to the tape-casting forming process. Graded-composition multilayer structures and microstructure are then elaborated by co-sintering. Performances in terms of membrane oxygen flows are presented. Mechanisms limiting the oxygen flow are discussed in order to propose ways of improving membrane performances. The limits of the studied system are defined in terms of elastic properties, and optimization ways are proposed for the dense membrane material composition in terms of mechanical properties and performance in oxygen semi-permeation

  7. Comparison of nuclear safety research reactor (TRIGA-ACPR) performance with analytical prediction

    International Nuclear Information System (INIS)

    West, G.B.; Whittemore, W.L.

    1976-01-01

    The NSRR was taken critical on June 30, 1975 at the Japan Atomic Energy Research Institute - Tokai-mura, Japan. Following initial core loading and control rod calibration, a series of pulsing tests was performed to characterize the performance of the reactor. A comparison has been made of performance parameters actually measured in the 157 element core versus predicted values based upon design analyses. The nuclear parameters measured were quite close to prediction. A $4.70 pulse produced a minimum period of 1.12 msec, a peak power of 20,500 MW and yielded a prompt energy release of 103 MW-sec. Pulse tests with experimental UO 2 fuel pins in the central irradiation cavity have produced 320 cal/gm, averaged at the axial center of 10% enriched UO 2 , for a 100 MW-sec pulse. The pulse rods for the NSRR contain B 4 C enriched to about 93 percent in Boron-10 in order to achieve maximum design performance with only three pulse rods. The total worth for the three transient rods was measured to be about $5.05 (vs $5.07 calculated for the 165 element core), thus verifying the effectiveness of the Boron-10 enrichment to achieve the desired result. Analysis of fuel temperature measurements made in the NSRR show that, for fuel temperatures produced during pulsing greater than 900 deg. C, heat transfer in the 0.010-inch gap between fuel and clad is enhanced by the minor outgassing of hydrogen which is characteristic of that temperature region. The hydrogen is normally all reabsorbed within about 100 sec of maximum temperature, at which time the heat transfer is characteristic of air (or argon) in the gap. In some of the temperature-instrumented elements, however, all of the hydrogen was not reabsorbed and as a result these elements gave significantly lower temperatures for high power steady state operation than were recorded prior to pulsing. In general, the NSRR parameters measured during startup were quite close to analytical prediction and the overall performance of the

  8. Application of expert system to evaluating reactor water cleanup system performance

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Nakamura, Masahiro; Nagasawa, Katsumi; Fushiki, Sumiyuki.

    1991-01-01

    Expert systems employing artificial intelligence (AI) have been developed for finding and elucidating causes of anomalies and malfunctions, presenting pertinent recommendation for countermeasures and for making precautionary diagnosis. On the other hand, further improvements in reliabilities for chemical control are required to promote BWR plant reliability and advancement. Especially, it is necessary to maintain the reactor water purity in high quality to minimize stress corrosion cracking (SCC) in primary cooling system, fuel performance degradation and radiation buildup. The reactor water quality is controlled by the reactor water cleanup (RWCU) system. So, it is very important to maintain the RWCU performance, in order to keep good reactor water quality. This paper describes an expert system used for evaluating RWCU system performance in BWR plants. (author)

  9. Outage performance improvement by state of the art reactor stud tensioning

    International Nuclear Information System (INIS)

    Oehler, Horst Werner; Vervliet, Herman

    2006-01-01

    Actual methods of reactor closing, i.e. cover to vessel sealing, is based on the creation of an equal load to the sealing circumference by tensioning all reactor studs with an equal force. This method ensures leak tightness through equal compression of the reactor seal in normal circumstances and is largely applied for all types of reactors throughout many generations and designs of nuclear power stations. The tension generated in each reactor stud is controlled indirectly by measuring the reactor stud elongation while under stress. Most studs are designed to measure this elongation easily by conventional or more advanced systems (from individual clock gauge to integrated digital transmission to a computer screen). It is this elongation value, prescribed by the reactor vessel/cover manufacturer which must be respected and demonstrated during all reactor closing operations, weather they take place for initial hydro testing, refuelling operations or periodical hydraulic tests of the primary circuit. Closing (and re-opening) of reactor vessels has become a routine operation as it is required for fuel reloading of the reactor core. This operation is performed on all PWR and BWR type of reactors with a large variety of tooling. As most of the utilities have implemented maintenance optimisation programs, the refuelling outage is reduced to a sequence of activities that allow quick and efficient refuelling of the core. The performance and efficiency of instrumentation and tooling deployed during these essential activities are of the utmost importance to minimise the critical path of the refuelling outage. Today, in support of outage performance, many utilities have invested in new and refurbished tooling to allow quick and efficient opening and closing of the reactor vessel. The features and properties of the most performing multi stud tensioning machines currently in service in nuclear power stations world wide (Africa, Europe, Asia and USA) are presented in the paper

  10. Performance analysis of Brayton cycle system for space power reactor

    International Nuclear Information System (INIS)

    Li Zhi; Yang Xiaoyong; Zhao Gang; Wang Jie; Zhang Zuoyi

    2017-01-01

    The closed Brayton cycle system now is the potential choice as the power conversion system for High Temperature Gas-cooled Reactors because of its high energy conversion efficiency and compact configuration. The helium is the best working fluid for the system for its chemical stability and small neutron absorption cross section. However, the Helium has small mole mass and big specific volume, which would lead to larger pipes and heat exchanger. What's more, the big compressor enthalpy rise of helium would also lead to an unacceptably large number of compressor's stage. For space use, it's more important to satisfy the limit of the system's volume and mass, instead of the requirement of the system's thermal capacity. So Noble-Gas binary mixture of helium and xenon is presented as the working fluid for space Brayton cycle. This paper makes a mathematical model for space Brayton cycle system by Fortran language, then analyzes the binary mixture of helium and xenon's properties and effects on power conversion units of the space power reactor, which would be helpful to understand and design the space power reactor. The results show that xenon would lead to a worse system's thermodynamic property, the cycle's efficiency and specific power decrease as xenon's mole fraction increasing. On the other hand, proper amount of xenon would decrease the enthalpy changes in turbomachines, which would be good for turbomachines' design. Another optimization method – the specific power optimization is also proposed to make a comparison. (author)

  11. Steam generator performance improvements for integral small modular reactors

    Directory of Open Access Journals (Sweden)

    Muhammad Ilyas

    2017-12-01

    Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure. The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

  12. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  13. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  14. Hospital ownership, decisions on supervisory board characteristics, and financial performance.

    Science.gov (United States)

    Kuntz, Ludwig; Pulm, Jannis; Wittland, Michael

    2016-01-01

    Dynamic and complex transformations in the hospital market increase the relevance of good corporate governance. However, hospital performance and the characteristics of supervisory boards differ depending on ownership. The question therefore arises whether hospital owners can influence performance by addressing supervisory board characteristics. The objective of this study is to explain differences in the financial performance of hospitals with regard to ownership by studying the size and composition of supervisory boards. The AMADEUS database was used to collect information on hospital financial performance in 2009 and 2010. Business and quality reports, hospital websites, and data from health insurer were used to obtain information on hospital and board characteristics. The resulting sample consisted of 175 German hospital corporations. We utilized ANOVA and regression analysis to test a mediation hypothesis that investigated whether decisions regarding board size and composition were associated with financial performance and could explain performance differences. Financial performance and board size and composition depend on ownership. An increase in board size and greater politician participation were negatively associated with all five tested measures of financial performance. Furthermore, an increase in physician participation was positively associated with one dimension of financial performance, whereas one negative relationship was identified for nurse and economist participation. For clerics, no associations were found. Decisions concerning board size and composition are important as they relate to hospital financial performance. We contribute to existing research by showing that, in addition to board size and physician participation, the participation of other professionals can also influence financial performance.

  15. Aggregate packing characteristics of good and poor performing asphalt mixes

    CSIR Research Space (South Africa)

    Denneman, E

    2007-07-01

    Full Text Available The aggregate structure of the compacted mix is a determining factor for the performance of Hot-Mix Asphalt (HMA). In this paper, the grading characteristics of good and poor performing HMA mixes are explored using the concepts of the Bailey method...

  16. Relationship Between Job Characteristics And Job Performance Of ...

    African Journals Online (AJOL)

    The agricultural extension agent is a key stakeholder in extension systems. The nature of their work is so important that it has overriding effect on their job performance. This study investigates the relationship between job characteristics and job performance of agricultural extension agents in Imo and Rivers States, Nigeria.

  17. Influence of course characteristics, student characteristics, and behavior in learning management systems on student performance

    OpenAIRE

    Conijn, Rianne; Kleingeld, Ad; Matzat, Uwe; Snijders, Chris; van Zaanen, Menno

    2016-01-01

    The use of learning management systems (LMS) in education make it possible to track students’ online behavior. This data can be used for educational data mining and learning analytics, for example, by predicting student performance. Although LMS data might contain useful predictors, course characteristics and student characteristics have shown to influence student performance as well. However, these different sets of features are rarely combined or compared. Therefore, in the current study we...

  18. Comparative performance of UASB and anaerobic hybrid reactors for the treatment of complex phenolic wastewater.

    Science.gov (United States)

    Ramakrishnan, Anushuya; Surampalli, Rao Y

    2012-11-01

    The performance of an upflow anaerobic sludge blanket (UASB) reactor and an anaerobic hybrid reactor (AHR) was investigated for the treatment of simulated coal wastewater containing toxic phenolics at different hydraulic retention times (0.75-0.33d). Fast start-up and granulation of biomass could be achieved in an AHR (45d) than UASB (58d) reactor. Reduction of HRT from 1.5 to 0.33d resulted in a decline in phenolics removal efficiency from 99% to 77% in AHR and 95% to 68% in UASB reactor respectively. AHR could withstand 2.5 times the selected phenolics loading compared to UASB reactor that could not withstand even 1.2 times the selected phenolics loading. Residence time distribution (RTD) study revealed a plug flow regime in the AHR and completely mixed regime in UASB reactor respectively. Energy economics of the reactors revealed that 12,159MJd(-1) more energy can be generated using AHR than UASB reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

  19. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Murata, Hiroyuki; Sawada, Kenichi; Inasaka, Fujio; Aya, Izuo; Shiozaki, Koki

    1999-01-01

    By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)

  20. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  1. Studies on characteristics of fluid dynamics in the coal liquefaction reactor; Sekitan ekika hanno tonai no ryudo tokusei

    Energy Technology Data Exchange (ETDEWEB)

    Sakawaki, K.; Nogami, Y.; Inokuchi, K. [Mitsui SRC Development Co. Ltd., Tokyo (Japan); Mochizuki, M.; Imada, K. [Nippon Steel Corp., Tokyo (Japan); Tachikawa, N.; Moki, T.; Ishikawa, I. [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1996-10-28

    To design the coal liquefaction reactor of large scale plant in future, it is important to understand characteristics of fluid dynamics within the coal liquefaction reactor. In this study, to measure the fluid dynamics of liquid phase within the coal liquefaction reactor operated under high temperature and high pressure coal liquefaction condition, neutron attenuating tracer (NAT) technique, one of the tracer test methods, was applied using 1 t/d coal treating PSU. The residence time of liquid phase within the reactor can be measured by utilizing property of neutron of being absorbed by materials. The tracer was injected at the inlets of first and third reactors, and the neutron was counted at each outlet. The concentration of tracer was derived from the discrete value, to determine the residence time distribution of liquid phase. The mean residence time of liquid phase in the single first reactor and in the total three reactors were prolonged under the severe operation conditions of liquefaction. The more severe the liquefaction operation condition was, the more active the mixing of liquid phase was in the first reactor. It was found that the progress of reaction was accelerated. 2 refs., 5 figs., 1 tab.

  2. Nonlinear Performance Characteristics of Flux-Switching PM Motors

    Directory of Open Access Journals (Sweden)

    E. Ilhan

    2013-01-01

    Full Text Available Nonlinear performance characteristics of 3-phase flux-switching permanent magnet motors (FSPM are overviewed. These machines show advantages of a robust rotor structure and a high energy density. Research on the FSPM is predominated by topics such as modeling and machine comparison, with little emphasis given on its performance and limits. Performance characteristics include phase flux linkage, phase torque, and phase inductance. In the paper, this analysis is done by a cross-correlation of rotor position and armature current. Due to the high amount of processed data, which cannot be handled analytically within an acceptable time period, a multistatic 2D finite element model (FEM is used. For generalization, the most commonly discussed FSPM topology, 12/10 FSPM, is chosen. Limitations on the motor performance due to the saturation are discussed on each characteristic. Additionally, a focused overview is given on energy conversion loops and dq-axes identification for the FSPM.

  3. Covariance of engineering management characteristics with engineering employee performance

    Science.gov (United States)

    Hesketh, Andrew Arthur

    1998-12-01

    As business in the 1990's grapples with the impact of continuous improvement and quality to meet market demands, there is an increased need to improve the leadership capabilities of our managers. Engineers have indicated desire for certain managerial characteristics in their leadership but there have been no studies completed that approached the problem of determining what managerial characteristics were best at improving employee performance. This study addressed the idea of identifying certain managerial characteristics that enhance employee performance. In the early 1990's, McDonnell Douglas Aerospace in St. Louis used a forced distribution system and allocated 35% of its employees into a "exceeds expectations" category and 60% into a "meets expectations" category. A twenty-question 5 point Likert scale survey on managerial capabilities was administered to a sample engineering population that also obtained their "expectations" category. A single factor ANOVA on the survey results determined a statistical difference between the "exceeds" and "meets" employees with four of the managerial capability questions. The "exceeds expectations" employee indicated that supervision did a better job of supporting subordinate development, clearly communicating performance expectations, and providing timely performance feedback when compared to the "meets expectations" employee. The "meets expectations" employee felt that their opinions, when different from their supervisor's, were more often ignored when compared to the "exceeds expectations" employee. These four questions relate to two specific managerial characteristics, "gaining (informal) authority and support" or "control" characteristic and "providing assistance and guidance" or "command" characteristic, that can be emphasized in managerial training programs.

  4. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    International Nuclear Information System (INIS)

    Iracane, Daniel; Bignan, Gilles; Lindbaeck, Jan-Erik; Blomgren, Jan

    2010-01-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  5. Current Status of the LIFE Fast Reactors Fuel Performance Codes

    International Nuclear Information System (INIS)

    Yacout, A.M.; Billone, M.C.

    2013-01-01

    The LIFE-4 (Rev. 1) code was calibrated and validated using data from (U,Pu)O2 mixed-oxide fuel pins and UO2 blanket rods which were irradiation tested under steady-state and transient conditions. – It integrates a broad material and fuel-pin irradiation database into a consistent framework for use and extrapolation of the database to reactor design applications. – The code is available and running on different computer platforms (UNIX & PC) – Detailed documentations of the code’s models, routines, calibration and validation data sets are available. LIFE-METAL code is based on LIFE4 with modifications to include key phenomena applicable to metallic fuel, and metallic fuel properties – Calibrated with large database from irradiations in EBR-II – Further effort for calibration and detailed documentation. Recent activities with the codes are related to reactor design studies and support of licensing efforts for 4S and KAERI SFR designs. Future activities are related to re-assessment of the codes calibration and validation and inclusion of models for advanced fuels (transmutation fuels)

  6. Numerical simulation of flow characteristics of lean jet to cross-flow in safety injection of reactor cooling system

    International Nuclear Information System (INIS)

    Wang Haijun; He Huining; Luo Yushan; Wang Weishu

    2011-01-01

    In the present work, a numerical simulation was performed to study the flow characteristics of lean jet to cross flow in a main tube in the safety injection of reactor cooling system. The influence scope and mixing characteristics of the confined lean jet in cross-flow were studied. It can be concluded that three basic flow regimes are marked, namely the attached lean jet, lift-off lean jet and impinging lean jet. The velocity ratio V R is the key factor in the flow state. The depth and region of jet to main flow are enhanced with the increase of the velocity ratio. The jet flow penetrates through the main flow with the increase of the velocity ratio. At higher velocity ratio, the jet flow strikes the main flow bottom and circumfluence happens in upriver of main flow. The vortex flow characteristics dominate the flow near region of jet to cross-flow and the mixture of jet to cross-flow. At different velocity ratio V R , the vortex grows from the same displacement, but the vortex type and the vortex is different. At higher velocity ratio, the vortex develops fleetly, wears off sharp and dies out sharp. The study is very important to the heat transfer experiments of cross-flow jet and thermal stress analysis in the designs of nuclear engineering. (authors)

  7. A performance evaluation of a microchannel reactor for the production of hydrogen from formic acid for electrochemical energy applications

    CSIR Research Space (South Africa)

    Ndlovu, IM

    2017-12-01

    Full Text Available An experimental evaluation of a microchannel reactor was completed to assess the reactor performance for the catalytic decomposition of vaporised formic acid (FA) for H2 production. Initially, X-ray powder diffraction (XRD), elemental mapping using...

  8. Selection method and device for reactor core performance calculation input indication

    International Nuclear Information System (INIS)

    Yuto, Yoshihiro.

    1994-01-01

    The position of a reactor core component on a reactor core map, which is previously designated and optionally changeable, is displayed by different colors on a CRT screen by using data of a data file incorporating results of a calculation for reactor core performance, such as incore thermal limit values. That is, an operator specifies the kind of the incore component to be sampled on a menu screen, to display the position of the incore component which satisfies a predetermined condition on the CRT screen by different colors in the form of a reactor core map. The position for the reactor core component displayed on the CRT screen by different colors is selected and designated on the screen by a touch panel, a mouse or a light pen, thereby automatically outputting detailed data of evaluation for the reactor core performance of the reactor core component at the indicated position. Retrieval of coordinates of fuel assemblies to be data sampled and input of the coordinates and demand for data sampling can be conducted at once by one menu screen. (N.H.)

  9. Characteristics and economy of the European reactor of pressurized water (EPR); Caracteristicas y economia del reactor europeo de agua a presion (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jov@nuclear.inin.mx

    2005-07-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  10. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  11. Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.

    1988-12-01

    The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

  12. Effect of different materials in the performance of solar reactors deployed in Jaiba, Minas Gerais state

    Energy Technology Data Exchange (ETDEWEB)

    Sartori, Marcia Aparecida; Soares, Antonio Alves; Soares, Adilson Rodrigues; Batista, Rafael Oliveira; Leite, Caio Vinicius [Universidade Federal de Vicosa (DEA/UFV), MG (Brazil). Dept. de Engenharia Agricola

    2008-07-01

    This study aimed to analyze the effect of different materials (masonry, butyl canvas and fiberglass) in the performance of solar reactors deployed in the city of Jaiba, Minas Gerais State. To do so, mini-stations to treat the domestic sewage were assembled. During the tests, samples of the effluent were collected upstream and downstream of the septic tank and the solar reactor. Fecal coliforms, BOD and COD were quantified in laboratory. The results indicated that the materials tested for construction of the reactor did not influence the solar disinfection of fecal coliforms. (author)

  13. Analysis on transient hydrodynamic characteristics of cavitation process for reactor coolant pump

    International Nuclear Information System (INIS)

    Wang Xiuli; Wang Peng; Yuan Shouqi; Zhu Rongsheng; Fu Qiang

    2014-01-01

    The reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX, and the corresponding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform. The results show that gas content presents the law of exponential function with the pressure reduction or time increase. In the cavitation primary condition, the pulsation frequency of head for the reactor coolant pump is mainly low frequency, and the main frequency of pressure pulsation is still rotation frequency while the effect of the pressure pulsation caused by cavitation on main frequency is not obvious. With the development of cavitation, the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency, secondary frequency and pulsating amplitude while the head pulsation frequency is given priority to low frequency pulse. Under serious cavitation condition, the head pulsation frequency is given priority to irregular changes of pulse high frequency, and also contains almost regular changes of low frequency. (authors)

  14. Core characteristics of fast reactor cycle with simple dry pyrochemical processing

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo

    2008-01-01

    Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO 2 spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle. (author)

  15. The Assessment Of High Temperature Reactor Fuel (Characteristics Of HTTR Fuel)

    International Nuclear Information System (INIS)

    Dewita, Erlan; Tuka, Veronica; Gunandjar

    1996-01-01

    HTTR is one of the reactor type with Helium coolant and outlet coolant temperature of 950 o C. One possibility of HTTR application is the coo generation of steam in high temperature and electric power for supply energy to industry in the future. Considering to the high operating temperature of HTTR, therefore it is needed the reactor fuel which have good mechanical, chemical and physical stability to the high temperature, and stable to the influence of fission fragment and neutron during irradiation. This assessment of the HTTR fuel characteristic based on the experiment data to find information of HTTR operation feasibility. Result of the assessment indicated that fission gas release at burn-up of 3.6 % FIMA which was the same as the maximum burn up in the HTTR design was fairly lower than the maximum release estimated in the design (5 x 10 - 4), which is R/B from the fuel fabricated by the prismatic block fuel method would be low (between 10 - 9 dan 10 - 8)

  16. Influence of geometrical and operational parameters on the performance of porous catalytic membrane reactors

    NARCIS (Netherlands)

    Aran, H.C.; Klooster, H.J.G.; Jani, J.M.; Wessling, Matthias; Lefferts, Leonardus; Lammertink, Rob G.H.

    2012-01-01

    In this study, porous membrane reactors with various characteristic length (inner diameter), controllable catalyst support thickness, active catalyst surface area and tunable wetting properties are described for heterogeneously catalyzed gas¿liquid¿solid (G¿L¿S) reactions. We developed porous

  17. Calculational investigations and analysis of characteristics of research reactor WWR-M as a source of neutrons for solution of scientific and applied tasks

    International Nuclear Information System (INIS)

    Vorona, P.M.; Razbudej, V.F.

    2010-01-01

    Calculational studies and analysis of the neutron fields of WWR-M research reactor of the Institute for Nuclear Research, National Academy of Sciences of Ukraine, as a basic nuclear facility for performing the fundamental and applied investigations and for experimentalindustrial production of radioisotope products for various spheres of application are carried out. The calculations are carried out by the method of statistic tests (Monte Carlo) applying the computer program MCNP-4C. The data on the spectra and the neutron flux density values at the 10 MW reactor power for all technological facilities designed for the works with neutrons: 19 vertical experimental channels for irradiation of specimens and 10 horizontal channels for beams extraction from the reactor are obtained. The effect of the neutron traps (water cavities) mounted in the core on the characteristics of the extracted from the reactor beams is demonstrated. Recommendations associated with optimization of the reactor core are adduced for amplification of its capabilities as a neutron source in experimental researches.

  18. The dynamic characteristics of HTGR (High Temperature Gas Cooled Reactor) system, (2)

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko; Ohta, Masao; Kawasaki, Hidenori

    1979-01-01

    The dynamic characteristics of a HTGR plant, which has two cooling loops, was investigated. The analytical model consists of the core with fuel sleeves, coolant channels and blocks, the upper and lower reflectors, the high and low temperature plenums, two double wall pipings, two intermediate heat exchangers and the secondary system. The key plant parameters for calculation were as follows: the core outlet gas temperature 1000 deg C, the reactor thermal output 50 MW, the flow rate of primary coolant gas 7.96 kg/sec-loop and the pressure of primary coolant gas 40 kg/cm 2 at the rated operating condition. The calculating parameters were fixed as follows: the time interval for core characteristic analysis 0.1 sec, the time interval for thermal characteristic analysis 5.0 sec, the number of division of fuel channels 130, and the number of division of an intermediate heat exchanger 200. The assumptions for making the model were evaluated especially for the power distribution in the core and the heat transmission coefficients in the core, the double wall piping and the intermediate heat exchangers. Concerning the analytical results, the self-control to the outer disturbance of reactivity and the plant dynamic behavior due to the change of flow rate of primary and secondary coolants, and the change of gas temperature of secondary coolant at the inlet of intermediate heat exchangers, are presented. (Nakai, Y.)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  20. Design requirements and performance requirements for reactor fuel recycle manipulator systems

    International Nuclear Information System (INIS)

    Grundmann, J.G.

    1975-01-01

    The development of a new generation of remote handling devices for remote production work in support of reactor fuel recycle systems is discussed. These devices require greater mobility, speed and visual capability than remote handling systems used in research activities. An upgraded manipulator system proposed for a High-Temperature Gas-Cooled Reactor fuel refabrication facility is described. Design and performance criteria for the manipulators, cranes, and TV cameras in the proposed system are enumerated

  1. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

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

    International Nuclear Information System (INIS)

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

    1957-01-01

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

  3. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

    International Nuclear Information System (INIS)

    Pathania, R.S.; Tatone, O.S.

    1979-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

  4. Investigation of bi-enzymatic reactor based on hybrid monolith with nanoparticles embedded and its proteolytic characteristics.

    Science.gov (United States)

    Shangguan, Lulu; Zhang, Lingyi; Xiong, Zhichao; Ren, Jun; Zhang, Runsheng; Gao, Fangyuan; Zhang, Weibing

    2015-04-03

    The bottom-up strategy of proteomic profiling study based on mass spectrometer (MS) has drawn high attention. However, conventional solution-based digestion could not satisfy the demands of highly efficient and complete high throughput proteolysis of complex samples. We proposed a novel bi-enzymatic reactor by immobilizing two different enzymes (trypsin/chymotrypsin) onto a mixed support of hybrid organic-inorganic monolith with SBA-15 nanoparticles embedded. Typsin and chymotrypsin were crossly immobilized onto the mixed support by covalent bonding onto the monolith with glutaraldehyde as bridge reagent and chelation via copper ion onto the nanoparticles, respectively. Compared with single enzymatic reactors, the bi-enzymatic reactor improved the overall functional analysis of membrane proteins of rat liver by doubling the number of identified peptides (from 1184/1010 with trypsin/chymotrypsin enzymatic reactors to 2891 with bi-enzymatic reactor), which led to more proteins identified with deep coverage (from 452/336 to 620); the efficiency of the bi-enzymatic reactor is also better than that of solution-based tandem digestion, greatly shorting the digestion time from 24h to 50s. Moreover, more transmembrane proteins were identified by bi-enzymatic reactor (106) compared with solution-based tandem digestion (95) with the same two enzymes and enzymatic reactors with single enzyme immobilized (75 with trypsin and 66 with chymotrypsin). The proteolytic characteristics of the bi-enzymatic reactors were evaluated by applying them to digestion of rat liver proteins. The reactors showed good digestion capability for proteins with different hydrophobicity and molecular weight. Copyright © 2015 Elsevier B.V. All rights reserved.

  5. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  6. Research of psychological characteristics and performance relativity of operators

    International Nuclear Information System (INIS)

    Fang Xiang; He Xuhong; Zhao Bingquan

    2008-01-01

    Based on the working tasks of an operator being taken into full consideration in this paper, on the one hand the table of measuring psychological characteristics is designed through the selection of special dimensions; on the other hand the table of performance appraisal is drafted through the choice of suitable standards of an operator. The paper analyzes the results of two aspects, sets relevant nuclear power plant operators as the research objective, and obtains the psychological characteristics and performance relativity of operators. The research can be as important and applied reference for the selection, evaluation and use of operators

  7. Performance and methanogenic community of rotating disk reactor packed with polyurethane during thermophilic anaerobic digestion

    International Nuclear Information System (INIS)

    Yang, Yingnan; Tsukahara, Kenichiro; Sawayama, Shigeki

    2007-01-01

    A newly developed anaerobic rotating disk reactor (ARDR) packed with polyurethane was used in continuous mode for organic waste removal under thermophilic (55 o C) anaerobic conditions. This paper reports the effects of the rotational speed on the methanogenic performance and community in an ARDR supplied with acetic acid synthetic wastewater as the organic substrate. The best performance was obtained from the ARDR with the rotational speed (ω) of 30 rpm. The average removal of dissolved organic carbon was 98.5%, and the methane production rate was 393 ml/l-reactor/day at an organic loading rate of 2.69 g/l-reactor/day. Under these operational conditions, the reactor had a greater biomass retention capacity and better reactor performance than those at other rotational speeds (0, 5 and 60 rpm). The results of 16S rRNA phylogenetic analysis indicated that the major methanogens in the reactor belonged to the genus Methanosarcina spp. The results of real-time polymerase chain reaction (PCR) analysis suggested that the cell density of methanogenic archaea immobilized on the polyurethane foam disk could be concentrated more than 2000 times relative to those in the original thermophilic sludge. Scanning electron microphotographs showed that there were more immobilized microbes at ω of 30 rpm than 60 rpm. A rotational speed on the outer layer of the disk of 6.6 m/min could be appropriate for anaerobic digestion using the polyurethane ARDR

  8. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

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

  9. Local flow distribution analysis inside the reactor pools of KALIMER-600 and PDRC performance test facility

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Hwang, Seong Won; Choi, Kyeong Sik

    2010-05-01

    In the study, 3-dimensional thermal hydraulic analysis was carried out focusing on the thermal hydraulic behavior inside the reactor pools for both KALIMER-600 and one-fifth scale-down test facility. STAR-CD, one of the commercial CFD codes, was used to analyze 3-dimensional incompressible steady-state thermal hydraulic behavior in both designs of KALIMER-600 and the scale-down test facility. In the KALIMER-600 CFD analysis, the pressure drops in the core and IHX gave a good agreement within 1% error range. It was found that the porous media model was appropriate to analyze the pressure distribution inside reactor core and IHX. Also, a validation analysis showed the pressure drop through the porous media under the condition of 80% flow rate and thermal power was calculated 64% less than in 100% condition giving a physically reasonable analytic result. Since the temperatures in the hot-side pool and cold-side pool were estimated to be very close to 540 and 390 .deg. C specified on the design values respectively, the CFD models of heat source and sink was confirmed. Through the study, the methodology of 3-dimensional CFD analysis about KALIMER-600 has been established and proven. Performed with the methodology, the analysis data such as flow velocity, temperature and pressure distribution were compared by normalizing those data for the actual sized modeling and scale-down modeling. As a result, the characteristics of thermal hydraulic behavior were almost identical for the actual sized modeling and scale-down modeling and the similarity scaling law used in the design of the sodium test facility by KAERI was found to be correct

  10. Reactor core T-H characteristics determination in case of parallel operation of different fuel assembly types

    International Nuclear Information System (INIS)

    Hermansky, J.; Petenyi, V.; Zavodsky, M.

    2009-01-01

    The WWER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60 deg symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results. (Authors)

  11. Simplified analysis of PRISM RVACS [Reactor Vessel Auxiliary Cooling System] performance without liner spill-over

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1990-01-01

    Simplified analysis of the performance of the PRISM RVACS decay heat removal system under off-normal conditions, i.e., without the liner spill-over, is described. Without the spilling of hot-pool sodium over the liner and the resultant down-flow along the inside of the reactor vessel wall, the RVACS system performance becomes dominated by the radial heat condition and radiation. Simple estimates of the resulting heat conduction and radiation processes support GE's contention that the RVACS performance is not severely impacted by the absence of spillover, and can improve significantly if sodium has leaked into the region between the reactor and containment vessels. 7 refs

  12. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  13. Facebook use, personality characteristics and academic performance: A correlational study

    OpenAIRE

    Sapsani, Georgia; Tselios, Nikolaos

    2017-01-01

    The present paper examines the relationship between the students personality, use of social media and their academic performance and engagement. In specific, the aim of this study is to examine the relationship of students facebook (fb) use and personality characteristics using the Big Five Personality Test with (a) student engagement, (b) time spent preparing for class, (c) time spent in co-curricular activities and (d) academic performance. Results illustrate that fb time was significantly ...

  14. Performance characteristics of the Mayo/IBM PACS

    Science.gov (United States)

    Persons, Kenneth R.; Gehring, Dale G.; Pavicic, Mark J.; Ding, Yingjai

    1991-07-01

    The Mayo Clinic and IBM (at Rochester, Minnesota) have jointly developed a picture archiving system for use with Mayo's MRI and Neuro CT imaging modalities. The communications backbone of the PACS is a portion of the Mayo institutional network: a series of 4-Mbps token rings interconnected by bridges and fiber optic extensions. The performance characteristics of this system are important to understand because they affect the response time a PACS user can expect, and the response time for non-PACS users competing for resources on the institutional network. The performance characteristics of each component and the average load levels of the network were measured for various load distributions. These data were used to quantify the response characteristics of the existing system and to tune a model developed by North Dakota State University Department of Computer Science for predicting response times of more complex topologies.

  15. Investigation of fuel lattice pitch changes influence on reactor performance through evaluate the neutronic parameters

    International Nuclear Information System (INIS)

    Zareian Ronizi, F.; Fadaei, A.H.; Setayeshi, S.; Shahidi, A.R.

    2015-01-01

    Highlights: • One of the most complex issues that Nu-engineers deal with is the design of NR core. • Numerous factors in nuclear core design depend on Fuel-to-Moderator volume ratio. • Aim of this research is to investigate RX performance for different lattice pitches. - Abstract: Nuclear reactor core design is one of the most complex issues that nuclear engineers deal with. The number and complexity of effective parameters and their impact on reactor design, which makes the problem difficult to solve, require precise knowledge of these parameters and their influence on the reactor operation. Numerous factors in a nuclear reactor core design depend on the Fuel-to-Moderator volume ratio, V F /V M , in a fuel cell. This ratio can be modified by changing the lattice pitch which is the thickness of water channels between fuels plates while keeping fuel slab dimensions fixed. Cooling and moderating properties of water are affected by such a change in a reactor core, and hence some parameters related to these properties might be changed. The aim of this research is to provide the suitable knowledge for nuclear core designing. To reach this goal, the first operating core of Tehran Research Reactor (TRR) with different lattice pitches is simulated, and the effect of different lattice pitches on some parameters such as effective multiplication factor (K eff ), reactor life time, distribution of neutron flux and power density in the core, as well as moderator temperature and density coefficient of reactivity are evaluated. The nuclear reactor analysis code, MTR-PC package is employed to carry out the considered calculation. Finally, the results are presented in some tables and graphs that provide useful information for nuclear engineers in the nuclear reactor core design

  16. Burnup performance of OTTO cycle pebble bed reactors with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor

  17. Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

    International Nuclear Information System (INIS)

    Hermansky, Jozef; Zavodsky, Martin

    2014-01-01

    A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

  18. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Tsige-Tamirat, H.; Ammirabile, L.; D' Agata, E.; Fuetterer, M.; Ranguelova, V. [European Commission, Joint Research Centre, Institute for Energy, Westerduinweg 3, 1755LE Petten (Netherlands)

    2010-07-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  19. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ammirabile, L.; D'Agata, E.; Fuetterer, M.; Ranguelova, V.

    2010-01-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  20. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  1. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  2. Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

    International Nuclear Information System (INIS)

    Sivaraman, N.; Subramaniam, S.; Srinivasan, T.G.; Vasudeva Rao, P.R.

    2002-01-01

    Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel. (author)

  3. Comparison of performance indicators of different types of reactors based on ISOE database

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2005-01-01

    The optimisation of the operation of a nuclear power plant (NPP) is a challenging issue due to the fact that besides general management issues, a risk associated to nuclear facilities should be included. In order to optimise the radiation protection programmes in around 440 reactors in operation with more than 500 000 monitored workers each year, the international exchange of performance indicators (PI) related to radiation protection issues seems to be essential. Those indicators are a function of a type of a reactor as well as the age and the quality of the management of the reactor. in general three main types of radiation protection PI could be recognised. These are: occupational exposure of workers, public exposure and management of PI related to radioactive waste. The occupational exposure could be efficiently studied using ISOC database. The dependence of occupational exposure on different types of reactors, e.g. PWR, BWR, are given, analysed and compared. (authors)

  4. Performance Characteristics of an Armature Voltage Controlled D.C. ...

    African Journals Online (AJOL)

    In this paper, the performance study of a separately excited d. c. motor whose speed is controlled by armature voltage variation is presented. Both the open loop and the closed loop steady state and transient characteristics are reported. The speed controllers considered in the closed loop mode are the proportional and the ...

  5. Performance and Carcass Characteristics of Broiler Finisher Birds ...

    African Journals Online (AJOL)

    Sixty (60) 4 weeks old Anak broiler strain were subjected to 28 days feeding trial at the Poultry Unit of the Teaching and Research Farm, Evan Enwerem, Owerri, Nigeria, to determine the dietary effect of pineapple wine sediment (PWSM) on their performance and carcass characteristics. The birds were divided into four ...

  6. Do the Managerial Characteristics of Schools Influence Their Performance?

    Science.gov (United States)

    Agasisti, Tommaso; Bonomi, Francesca; Sibiano, Piergiacomo

    2012-01-01

    Purpose: The purpose of this paper is to investigate the role of governance and managerial characteristics of schools. More specifically, the aim is to individuate the factors that are associated to higher schools' performances, as measured through student achievement. Design/methodology/approach: The research is conducted by means of a survey in…

  7. Performance characteristics of proximity focused ultraviolet image converters

    Science.gov (United States)

    Williams, J. T.; Feibelman, W. A.

    1973-01-01

    Performance characteristics of Bendix type BX 8025-4522 proximity focused image tubes for UV to visible light conversion are presented. Quantum efficiency, resolution, background, geometric distortion, and environmental test results are discussed. The converters use magnesium fluoride input windows with Cs-Te photocathodes and P-11 phosphors on fiber optic output windows.

  8. Growth performance, carcass and organ characteristics of growing ...

    African Journals Online (AJOL)

    An experiment was conducted at the Department of Animal Science teaching and research farm, Bayero University Kano, to evaluate the effect of feeding graded levels of Moringa oleifera leaf meal (MOLM) in diets on growth performance, carcass and organ characteristics of weaned rabbits. Twenty eight grower rabbits of ...

  9. Performance and ileal characteristics of finishing broilers fed diets ...

    African Journals Online (AJOL)

    An experiment was conducted to evaluate the effect of prebiotics supplemented diets on performance characteristics and gut morphology of broiler chickens. The study involved 320 day-old Anak broiler chicks, used to assess the utilization of prebiotics [Mannose oligosaccharides (MOS) and Lactose oligosaccharides ...

  10. Nonlinear performance characteristics of flux-switching PM motors

    NARCIS (Netherlands)

    Ilhan, E.; Kremers, M.F.J.; Motoasca, T.E.; Paulides, J.J.H.; Lomonova, E.

    2013-01-01

    Nonlinear performance characteristics of 3-phase flux-switching permanent magnet motors (FSPM) are overviewed. These machines show advantages of a robust rotor structure and a high energy density. Research on the FSPM is predominated by topics such as modeling and machine comparison, with little

  11. Performance characteristics of broiler chicks fed kidney bean as ...

    African Journals Online (AJOL)

    An experiment was conducted to investigate the effect of replacing soybean meal and groundnut cake meal with cooked and decorticated kidney bean seed meals on the performance characteristics of broilers. One hundred and eighty day old broiler chicks of Anak strain were raised on six experimental diets.

  12. An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1995-07-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel

  13. Design characteristics of zero power fast reactor Lasta; Osnovne karakteristike brzog reaktora nulte snage Lasta

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Stefanovic, D; Pesic, M; Popovic, D; Nikolic, D; Antic, D; Zavaljevski, N [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)

    1987-07-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  14. Characteristics of Soil Structure Interaction for Reactor Building of Kashiwazaki-Kariwa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gil, Moon Joo; Jung, Rae Young; Hyun, Chang Hun; Kim, Moon Soo; Lim, Nam Hyoung

    2010-01-01

    On 16 July 2007, the Nigataken-chuetsu-oki earthquake registering a moment magnitude of 6.8 occurred at a depth of about 15 km. As a result of this earthquake, noticeable shaking exceeding the design ground motion was measured at the Tokyo Electric Power Company (TEPCO) Kashiwazaki-Kariwa Nuclear Power Station (KKN), the biggest nuclear power plant in the world, located at about 16 km away from the epicenter. This earthquake triggered a fire at an electrical transformer and insignificant damage on some parts of facilities. This event gave an impulse to study on the damage and safety margin of nuclear power plant due to the strong earthquake exceeding design basis. As a part of those efforts, KARISMA (KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment) benchmark study was launched by the IAEA in terms of an international collaborative research. The main objectives of this research are to estimate the structural behavior and to evaluate the seismic margin of reactor building considering the effects of Soil-Structure Interaction (SSI). This paper presents verification of structural model developed here and validation of soil foundation characteristics through soil-column analysis. It has also been demonstrated that the spring constants and damping coefficient obtained from impedance analysis represent well the soil foundation characteristics

  15. A review of boiling water reactor water chemistry: Science, technology, and performance

    International Nuclear Information System (INIS)

    Fox, M.J.

    1989-02-01

    Boiling water reactor (BWR) water chemistry (science, technology, and performance) has been reviewed with an emphasis on the relationships between BWR water quality and corrosion fuel performance, and radiation buildup. A comparison of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.56, the Boiling Water Reactor Owners Group (BWROG) Water Chemistry Guidelines, and Plant Technical Specifications showed that the BWROG Guidelines are more stringent than the NRC Regulatory Guide, which is almost identical to Plant Technical Specifications. Plant performance with respect to BWR water chemistry has shown dramatic improvements in recent years. Up until 1979 BWRs experienced an average of 3.0 water chemistry incidents per reactor-year. Since 1979 the water chemistry technical specifications have been violated an average of only 0.2 times per reactor-year, with the most recent data from 1986-1987 showing only 0.05 violations per reactor-year. The data clearly demonstrate the industry-wide commitment to improving water quality in BWRs. In addition to improving water quality, domestic BWRs are beginning to switch to hydrogen water chemistry (HWC), a remedy for intergranular stress corrosion cracking. Three domestic BWRs are presently operating on HWC, and fourteen more have either performed HWC mini tests or are in various stages of HWC implementation. This report includes a detailed review of HWC science and technology as well as areas in which further research on BWR chemistry may be needed. 43 refs., 30 figs., 8 tabs

  16. Welding and cutting characteristics of blanket/first wall module to back plate for fusion experimental reactor

    International Nuclear Information System (INIS)

    Sato, Shinichi; Osaki, Toshio; Koga, Shinji

    1996-01-01

    The first wall and the blanket of the International Thermonuclear Experimental Reactor (ITER) are used under severe conditions such as the neutron irradiation by plasma, surface thermal load, the electromagnetic force at the time of plasma disruption and others. Consequently, from the viewpoint of the necessity for disassembling and maintenance, those are divided into modules in toroidal and poloidal directions. In this study, as to the welding of the back plate and the legs supporting blanket modules, which are installed in a vacuum vessel, the characteristic test paying attention to the deformation at the time of welding was carried out, and the optimal welding conditions and the characteristics of welding deformation and others were clarified. Moreover, when water jet method was used for cutting the welded parts of the supporting legs, the properties of the cut parts, the time for cutting and others were examined. The performance required for the welded parts of blanket modules with back plate is shown. The basic test of welding conditions using plate models, partial model test and whole model test are reported. The test of water jet cutting for the maintenance of shielding blanket modules is described. (K.I.)

  17. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  18. A review of hospital characteristics associated with improved performance.

    Science.gov (United States)

    Brand, Caroline A; Barker, Anna L; Morello, Renata T; Vitale, Michael R; Evans, Sue M; Scott, Ian A; Stoelwinder, Johannes U; Cameron, Peter A

    2012-10-01

    The objective of this review was to critically appraise the literature relating to associations between high-level structural and operational hospital characteristics and improved performance. The Cochrane Library, MEDLINE (Ovid), CINAHL, proQuest and PsychINFO were searched for articles published between January 1996 and May 2010. Reference lists of included articles were reviewed and key journals were hand searched for relevant articles. and data extraction Studies were included if they were systematic reviews or meta-analyses, randomized controlled trials, controlled before and after studies or observational studies (cohort and cross-sectional) that were multicentre, comparative performance studies. Two reviewers independently extracted data, assigned grades of evidence according to the Australian National Health and Medical Research Council guidelines and critically appraised the included articles. Data synthesis Fifty-seven studies were reported within 12 systematic reviews and 47 observational articles. There was heterogeneity in use and definition of performance outcomes. Hospital characteristics investigated were environment (incentives, market characteristics), structure (network membership, ownership, teaching status, geographical setting, service size) and operational design (innovativeness, leadership, organizational culture, public reporting and patient safety practices, information technology systems and decision support, service activity and planning, workforce design, staff training and education). The strongest evidence for an association with overall performance was identified for computerized physician order entry systems. Some evidence supported the associations with workforce design, use of financial incentives, nursing leadership and hospital volume. There is limited, mainly low-quality evidence, supporting the associations between hospital characteristics and healthcare performance. Further characteristic-specific systematic reviews are

  19. Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

    2014-07-01

    Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

  20. Licensing process characteristics of Small Modular Reactors and spent nuclear fuel repository

    Energy Technology Data Exchange (ETDEWEB)

    Söderholm, Kristiina, E-mail: kristiina.soderholm@fortum.com [Fortum Power (Finland); Tuunanen, Jari, E-mail: jari.tuunanen@fortum.com [Fortum Power (Finland); Amaba, Ben, E-mail: baamaba@us.ibm.com [IBM Complex Systems (United States); Bergqvist, Sofia, E-mail: sofia.bergqvist@se.ibm.com [IBM Rational Software (Sweden); Lusardi, Paul, E-mail: plusardi@nuscalepower.com [NuScale Power (United States)

    2014-09-15

    Highlights: • We examine the licensing process challenges of modular nuclear facilities. • We compare the features of Small Modular Reactors and spent nuclear fuel repository. • We present the need of nuclear licensing simplification. • Part of the licensing is proposed to be internationally applicable. • Systems engineering and requirements engineering benefits are presented. - Abstract: This paper aims to increase the understanding of the licensing processes characteristics of Small Modular Reactors (SMR) compared with licensing of spent nuclear fuel repository. The basis of the SMR licensing process development lies in licensing processes used in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their various licensing processes and recent actions in the new NPP construction. Certain aspects of the aviation industry licensing process have also been studied and selected practices have been investigated as possibly suitable for use in nuclear licensing. Suitable features for SMR licensing are emphasized and suggested. The licensing features of the spent nuclear fuel deep repository along with similar features of SMR licensing are discussed. Since there are similar types of challenges of lengthy licensing time frames, as well as modular features to be taken into account in licensing, these two different nuclear industry fields can be compared. The main SMR features to take into account in licensing are: • Standardization of the design. • Modularity. • Mass production. • Serial construction. Modularity can be divided into two different categories: the first category is simply a single power plant unit constructed of independently engineered modules (e.g. construction process for Westinghouse AP-1000 NPP) and the second one a power plant composed of many reactor modules, which are manufactured in factories and installed as needed (e.g. NuScale Power SMR design). The deep underground repository

  1. Relations between mental health team characteristics and work role performance.

    Science.gov (United States)

    Fleury, Marie-Josée; Grenier, Guy; Bamvita, Jean-Marie; Farand, Lambert

    2017-01-01

    Effective mental health care requires a high performing, interprofessional team. Among 79 mental health teams in Quebec (Canada), this exploratory study aims to 1) determine the association between work role performance and a wide range of variables related to team effectiveness according to the literature, and to 2) using structural equation modelling, assess the covariance between each of these variables as well as the correlation with other exogenous variables. Work role performance was measured with an adapted version of a work role questionnaire. Various independent variables including team manager characteristics, user characteristics, team profiles, clinical activities, organizational culture, network integration strategies and frequency/satisfaction of interactions with other teams or services were analyzed under the structural equation model. The later provided a good fit with the data. Frequent use of standardized procedures and evaluation tools (e.g. screening and assessment tools for mental health disorders) and team manager seniority exerted the most direct effect on work role performance. While network integration strategies had little effect on work role performance, there was a high covariance between this variable and those directly affecting work role performance among mental health teams. The results suggest that the mental healthcare system should apply standardized procedures and evaluation tools and, to a lesser extent, clinical approaches to improve work role performance in mental health teams. Overall, a more systematic implementation of network integration strategies may contribute to improved work role performance in mental health care.

  2. Impact of asymmetric lamp positioning on the performance of a closed-conduit UV reactor

    Directory of Open Access Journals (Sweden)

    Tipu Sultan

    2017-06-01

    Full Text Available Computational fluid dynamics (CFD analyses for the performance improvement of a closed-conduit ultraviolet (UV reactor were performed by changing the lamp positions from symmetric to asymmetric. The asymmetric lamp positioning can be useful for UV reactor design and optimization. This goal was achieved by incorporating the two performance factors, namely reduction equivalent dose (RED and system dose performance. Four cases were carried out for asymmetric lamp positioning within the UV reactor chamber and each case consisted of four UV lamps that were simulated once symmetrically and four times asymmetrically. The results of the four asymmetric cases were compared with the symmetric one. Moreover, these results were evaluated by using CFD simulations of a closed-conduit UV reactor. The fluence rate model, UVCalc3D was employed to validate the simulations results. The simulation results provide detailed information about the dose distribution, pathogen track modeling and RED. The RED value was increased by approximately 15% by using UVCalc3D fluence rate model. Additionally, the asymmetric lamp positioning of the UV lamps had more than 50% of the pathogens received a better and a higher UV dose than in the symmetric case. Consequently, the system dose performance was improved by asymmetric lamp positioning. It was concluded that the performance parameters (higher RED and system dose performance were improved by using asymmetric lamp positioning.

  3. Decaffeination process characteristic of Robusta coffee in single column reactor using ethyl acetate solvent

    Directory of Open Access Journals (Sweden)

    Sukrisno Widyotomo

    2009-08-01

    Full Text Available Consumers drink coffee not as nutrition source, but as refreshment drink. For coffee consumers who have high tolerance for caffeine, coffee may warm up and refresh their bodies. High caffeine content in coffee beans may cause several complaints to consumers who are susceptible to caffeine. One of the efforts, for coffee market expansion is product diversification to decaffeinated coffee. Decaffeination process is one of process to reduce caffeine content from agricultural products. Indonesian Coffee and Cocoa Research Institute in collaboration with Bogor Agricultural University has developed a single column reactor for coffee beans decaffeination. The aim of this research is to study process characteristic of coffee decaffeination in single column reactor using ethyl acetate (C4H8O2 solvent. Treatments applicated in the research were time and temperature process. Temperature treatment were 50—60OC, 60—70OC, 70—80OC, 80—90OC and 90—100OC. Time treatment were 2 h, 4 h, 6 h, 8 h, 10 h, and 12 h Size of Robusta coffee beans used were less than 5.5 mm (A4, between 5.5 mm and 6.5 mm (A3, between 6.5 mm and 7.5 mm (A2, and more than 7.5 mm (A1. The result showed that decaffeination process with ethyl acetate solvent will be faster when its temperature was higher and smaller bean size. For bean size less than 5,5 mm, decaffeination process by 10% ethyl acetat can be done 8—10 hours in 90—100OC solvent temperature or 12 hours in 60—70OC solvent temperature for 0.3% caffein content. Organoleptic test showed that 90—100OC temperature solvent treatment decreased coffee flavor, which aroma, bitterness and body values were 1.9 each . Key words : Coffee, caffeine, decaffeination, quality, single column.

  4. Job characteristics, flow, and performance: the moderating role of conscientiousness.

    Science.gov (United States)

    Demerouti, Evangelia

    2006-07-01

    The present article aims to show the importance of positive work-related experiences within occupational health psychology by examining the relationship between flow at work (i.e., absorption, work enjoyment, and intrinsic work motivation) and job performance. On the basis of the literature, it was hypothesized that (a) motivating job characteristics are positively related to flow at work and (b) conscientiousness moderates the relationship between flow and other ratings of (in-role and out-of-role) performance. The hypotheses were tested on a sample of 113 employees from several occupations. Results of moderated structural equation modeling analyses generally supported the hypotheses. Motivating job characteristics were predictive of flow, and flow predicted in-role and extra-role performance, for only conscientious employees.

  5. Parametric study of geohydrologic performance characteristics for geologic waste repositories

    International Nuclear Information System (INIS)

    Bailey, C.E.; Marine, I.W.

    1980-11-01

    One of the major objectives of the National Waste Terminal Storage Program is to identify potential geologic sites for storage and isolation of radioactive waste (and possibly irradiated fuel). Potential sites for the storage and isolation of radioactive waste or spent fuel in a geologic rock unit are being carefully evaluated to ensure that radionuclides from the stored waste or fuel will never appear in the biosphere in amounts that would constitute a hazard to the health and safety of the public. The objective of this report is to quantify and present in graphical form the effects of significant geohydrologic and other performance characteristics that would influence the movement of radionuclides from a storage site in a rock unit to the biosphere. The effort in this study was focused on transport by groundwater because that is the most likely method of radionuclide escape. Graphs of the major performance characteristics that influence the transport of radionuclides from a repository to the biosphere by groundwater are presented. The major characteristics addressed are radioactive decay, leach rate, hydraulic conductivity, porosity, groundwater gradient, hydrodynamic dispersion, ion exchange, and distance to the biosphere. These major performance characteristics are combind with each other and with the results of certain other combinations and presented in graphical form to provide the interrelationships of values measured during field studies. The graphical form of presentation should be useful in the screening process of site selection. An appendix illustrates the use of these graphs to assess the suitability of a site

  6. Performance analyses of the communication networks of a modern supervision and control system of research reactors

    International Nuclear Information System (INIS)

    El-Madbouly, E.I.; Shaat, M.K.; Shokr, A.M.; Elrefaei, G.H.

    2009-01-01

    The functions of the Instrumentation and Control (I and C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated. (orig.)

  7. Performance analyses of the communication networks of a modern supervision and control system of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    El-Madbouly, E.I. [Menoufia Univ., Menouf (Egypt). Faculty of Electronics Engineering; Shaat, M.K.; Shokr, A.M.; Elrefaei, G.H. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor

    2009-04-15

    The functions of the Instrumentation and Control (I and C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated. (orig.)

  8. General performance characteristics of an irreversible ferromagnetic Stirling refrigeration cycle

    International Nuclear Information System (INIS)

    Lin, G.; Tegus, O.; Zhang, L.; Brueck, E.

    2004-01-01

    A new magnetic-refrigeration-cycle model using ferromagnetic materials as a cyclic working substance is set up, in which finite-rate heat transfer, heat leak and regeneration time are taken into account. On the basis of the thermodynamic properties of a ferromagnetic material, the general performance characteristics of the ferromagnetic Stirling refrigeration cycle are investigated and the effects of some key irreversibilities on the performance of the cycle are revealed. By using the optimal-control theory, the optimal relation between the coefficient of performance and the cooling rate is derived and some important performance bounds, e.g., the maximum cooling rate, the maximum coefficient of performance, are determined. Moreover, the optimal operating regions for cooling rate, coefficient of performance and the optimal operating temperatures of a cyclic working substance in the two heat-transfer processes are obtained. Furthermore, the influences of magnetization and magnetic field on the performance characteristics of the cycle are discussed. The results obtained here have general significance and can be deduced to the related ones of the Stirling refrigeration cycle using paramagnetic salt as a cyclic working substance

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

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

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

  10. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data

  11. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  12. Electron versus proton accelerator driven sub-critical system performance using TRIGA reactors at power

    International Nuclear Information System (INIS)

    Carta, M.; Burgio, N.; D'Angelo, A.; Santagata, A.; Petrovich, C.; Schikorr, M.; Beller, D.; Felice, L. S.; Imel, G.; Salvatores, M.

    2006-01-01

    This paper provides a comparison of the performance of an electron accelerator-driven experiment, under discussion within the Reactor Accelerator Coupling Experiments (RACE) Project, being conducted within the U.S. Dept. of Energy's Advanced Fuel Cycle Initiative (AFCI), and of the proton-driven experiment TRADE (TRIGA Accelerator Driven Experiment) originally planned at ENEA-Casaccia in Italy. Both experiments foresee the coupling to sub-critical TRIGA core configurations, and are aimed to investigate the relevant kinetic and dynamic accelerator-driven systems (ADS) core behavior characteristics in the presence of thermal reactivity feedback effects. TRADE was based on the coupling of an upgraded proton cyclotron, producing neutrons via spallation reactions on a tantalum (Ta) target, with the core driven at a maximum power around 200 kW. RACE is based on the coupling of an Electron Linac accelerator, producing neutrons via photoneutron reactions on a tungsten-copper (W-Cu) or uranium (U) target, with the core driven at a maximum power around 50 kW. The paper is focused on analysis of expected dynamic power response of the RACE core following reactivity and/or source transients. TRADE and RACE target-core power coupling coefficients are compared and discussed. (authors)

  13. Computationally-generated nuclear forensic characteristics of early production reactors with an emphasis on sensitivity and uncertainty

    International Nuclear Information System (INIS)

    Redd, Evan M.; Sjoden, Glenn; Erickson, Anna

    2017-01-01

    Highlights: •X-10 reactor is used as a case study for nuclear forensic signatures. •S/U analysis is conducted to derive statistically relevant markers. •Computationally-generated signatures aid with proliferation pathway identification. •Highest uncertainty in total plutonium production originates from 238 Pu and 242 Pu. -- Abstract: With nuclear technology and analysis advancements, site access restrictions, and ban on nuclear testing, computationally-generated nuclear forensic signatures are becoming more important in gaining knowledge to a reclusive country’s weapon material production capabilities. In particular, graphite-moderated reactors provide an appropriate case study for isotopics relevant in Pu production in a clandestine nuclear program due to the ease of design and low thermal output. We study the production characteristics of the X-10 reactor with a goal to develop statistically-relevant nuclear forensic signatures from early Pu production. In X-10 reactor, a flat flux gradient and low burnup produce exceptionally pure Pu as evident by the 240 Pu/ 239 Pu ratio. However, these design aspects also make determining reactor zone attribution, done with the 242 Pu/ 240 Pu ratio, uncertain. Alternatively, the same ratios produce statistically differentiable results between Manhattan Project and post-Manhattan Project reactor configurations, allowing for attribution conclusions.

  14. Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

    International Nuclear Information System (INIS)

    Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

    2006-01-01

    This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

  15. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  16. Reviewing real-time performance of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Preckshot, G.G. [Lawrence Livermore National Lab., CA (United States)

    1993-08-01

    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

  17. Reviewing real-time performance of nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Preckshot, G.G.

    1993-08-01

    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems

  18. Studies of severe accidents in light water reactors. Containment performance

    International Nuclear Information System (INIS)

    Hayns, M.R.; Phillips, D.W.; Young, R.L.D.

    1987-01-01

    The containment system of a LWR is an obvious component of the plant which performs an important safety function in preventing the release of fission products to the environment in the event of design basis accidents. With over 260 LWRs in service worldwide, and others still under construction, there is a considerable diversity of containment types and combinations of containment safeguards systems. All of these satisfy local regulatory requirements which are principally aimed at the design basis accidents, and these requirements naturally have a considerable uniformity. However, their design diversity becomes more relevant to the performance of the containment in severe accident conditions, and this aspect of containment performance is reviewed in this paper. The ability of the containment to mitigate severe accident consequences introduces the potential for accident management and recovery and this in turn points towards a range of new containment systems and concepts. PSA helps in judging these possibilities and in forming policies and procedures for accident management. It is perhaps in accident management that severe accident containment performance will be most beneficial in the future, and where additional effort in containment analysis will be focused

  19. Multidimensional performance characteristics and standard of performance in talented youth field hockey players : A longitudinal study

    NARCIS (Netherlands)

    Elferink-Gemser, Marije T.; Visscher, Chris; Lemmink, Koen A. P. M.; Mulder, Theo

    2007-01-01

    To identify performance characteristics that could help predict future elite field hockey players, we measured the anthropometric, physiological, technical, tactical, and psychological characteristics of 30 elite and 35 sub-elite youth players at the end of three consecutive seasons. The mean age of

  20. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    International Nuclear Information System (INIS)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

  1. Influence of prey body characteristics and performance on predator selection.

    Science.gov (United States)

    Holmes, Thomas H; McCormick, Mark I

    2009-03-01

    At the time of settlement to the reef environment, coral reef fishes differ in a number of characteristics that may influence their survival during a predatory encounter. This study investigated the selective nature of predation by both a multi-species predator pool, and a single common predator (Pseudochromis fuscus), on the reef fish, Pomacentrus amboinensis. The study focused on the early post-settlement period of P. amboinensis, when mortality, and hence selection, is known to be highest. Correlations between nine different measures of body condition/performance were examined at the time of settlement, in order to elucidate the relationships between different traits. Single-predator (P. fuscus) choice trials were conducted in 57.4-l aquaria with respect to three different prey characteristics [standard length (SL), body weight and burst swimming speed], whilst multi-species trials were conducted on open patch reefs, manipulating prey body weight only. Relationships between the nine measures of condition/performance were generally poor, with the strongest correlations occurring between the morphological measures and within the performance measures. During aquaria trials, P. fuscus was found to be selective with respect to prey SL only, with larger individuals being selected significantly more often. Multi-species predator communities, however, were selective with respect to prey body weight, with heavier individuals being selected significantly more often than their lighter counterparts. Our results suggest that under controlled conditions, body length may be the most important prey characteristic influencing prey survival during predatory encounters with P. fuscus. In such cases, larger prey size may actually be a distinct disadvantage to survival. However, these relationships appear to be more complex under natural conditions, where the expression of prey characteristics, the selectivity fields of a number of different predators, their relative abundance, and

  2. RELATIONSHIPS BETWEEN MUSCLE FATIGUE CHARACTERISTICS AND MARKERS OF ENDURANCE PERFORMANCE

    Directory of Open Access Journals (Sweden)

    Martyn G. Morris

    2008-12-01

    Full Text Available The aim of this study was to examine the relationship of a range of in-vivo whole muscle characteristics to determinants of endurance performance. Eleven healthy males completed a cycle ergometer step test to exhaustion for the determination of the lactate threshold, gross mechanical efficiency, peak power and VO2max. On two separate occasions, contractile and fatigue characteristics of the quadriceps femoris were collected using a specially designed isometric strength-testing chair. Muscle fatigue was then assessed by stimulating the muscle for 3 minutes. Force, rate of force development and rates of relaxation were calculated at the beginning and end of the 3 minute protocol and examined for reliability and in relation to lactate threshold, VO2max, gross mechanical efficiency and peak power. Muscle characteristics, rate of force development and relaxation rate were demonstrated to be reliable measures. Force drop off over the 3 minutes (fatigue index was related to lactate threshold (r = -0.72 p < 0.01 but not to VO2max. The rate of force development related to the peak power at the end of the cycle ergometer test (r = -0.75 p < 0.01. Rates of relaxation did not relate to any of the performance markers. We found in-vivo whole muscle characteristics, such as the fatigue index and rate of force development, relate to specific markers of peripheral, but not to central, fitness components. Our investigation suggests that muscle characteristics assessed in this way is reliable and could be feasibly utilised to further our understanding of the peripheral factors underpinning performance

  3. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  4. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

  5. NF-6 program complex for BESM-6 computation of the basic neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Zizin, M.N.; Savochkina, O.A.; Chukhlova, O.P.

    1978-01-01

    A structure of standard designations is described and semantics of a number of standard values used in a NF-6 program complex is given. Main source data and results of neutron-physical reactor calculation are standard values, the peculiarities of FORTRAN and ALGOL-GDR algorithm languages in the DUBNA monitoring system were taken account of. As a base of standard values list the FIHAR system list, supplemented with new standard designations for integral reactor characteristics, is used. Developed is also a list of standard values to organize the exchange with external memory in the process of task solution and long-range storage

  6. Preliminary investigation of fuel cycle in fast reactors by the correlations method and sensitivity analyses of nuclear characteristics

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Castro Lobo, P.D. de.

    1980-11-01

    A reduction of computing effort was achieved as a result of the application of space - independent continuous slowing down theory in the spectrum averaged cross sections and further expressing then in a quadratic corelation whith the temperature and the composition. The decoupling between variables that express some of the important nuclear characteristics allowed to introduce a sensitivity analyses treatment for the full prediction of the behavior, over the fuel cycle, of the LMFBR considered. As a potential application of the method here in developed is to predict the nuclear characteristics of another reactor, face some reference reactor of the family considered. Excellent agreement with exact calculation is observed only when perturbations occur in nuclear data and/or fuel isotopic characteristics, but fair results are obtained whith variations in system components other than the fuel. (Author) [pt

  7. Reactor containment purge and vent valve performance experiments

    International Nuclear Information System (INIS)

    Hunter, J.A.; Steele, R.; Watkins, J.C.

    1985-01-01

    Three nuclear-designed butterfly valves typical of those used in domestic nuclear power plant containment purge and vent applications were tested. For a comparison of responses, two eight-inch nominal pipe size valves with differing internal design were tested. For extrapolation insights, a 24-inch nominal pipe size valve was also tested. The valve experiments were performed with various piping configurations and valve disc orientations to the flow, to simulate various installation options in field application. As a standard for comparing the effects of the installation options, testing was also performed in a standard ANSI test section. Test cycles were performed at inlet pressures of 5 to 60 psig, while monitoring numerous test parameters, such as the valve disc position, valve shaft torque, mass flow rate, and the pressure and temperature at multiple locations throughout the test section. An experimental data base was developed to assist in the evaluation of the current analytical methods and to determine the influence of inlet pressure, inlet duct geometry, and valve orientation to the flow media on valve torque requirements, along with any resulting limitations to the extrapolation methods. 2 refs., 15 figs

  8. Relationships between Isometric Force-Time Characteristics and Dynamic Performance

    Directory of Open Access Journals (Sweden)

    Thomas Dos’Santos

    2017-09-01

    Full Text Available The purpose of this study was to explore the relationships between isometric mid-thigh pull (IMTP force-time characteristics (peak force and time-specific force vales (100–250 ms and dynamic performance and compare dynamic performance between stronger and weaker athletes. Forty-three athletes from different sports (rowing, soccer, bicycle motocross, and hockey performed three trials of the squat jump (SJ, countermovement jump (CMJ, and IMTP, and performed a one repetition maximum power clean (PC. Reactive strength index modified (RSImod was also calculated from the CMJ. Statistically significant large correlations between IMTP force-time characteristics and PC (ρ = 0.569–0.674, p < 0.001, and moderate correlations between IMTP force-time characteristics (excluding force at 100 ms and RSImod (ρ = 0.389–0.449, p = 0.013–0.050 were observed. Only force at 250 ms demonstrated a statistically significant moderate correlation with CMJ height (ρ = 0.346, p = 0.016 and no statistically significant associations were observed between IMTP force-time characteristics and SJ height. Stronger athletes (top 10 demonstrated statistically significantly greater CMJ heights, RSImods, and PCs (p ≤ 0.004, g = 1.32–1.89 compared to weaker (bottom 10 athletes, but no differences in SJ height were observed (p = 0.871, g = 0.06. These findings highlight that the ability to apply rapidly high levels of force in short time intervals is integral for PC, CMJ height, and reactive strength.

  9. EXPERIMENTAL ANALYSIS OF THE CHARACTERISTIC PERFORMANCE OF STANDALONE PHOTOVOLTAIC SYSTEM

    OpenAIRE

    Birendra Kishore; Anirban Nandy*; O.P. Pandey

    2016-01-01

    This paper demonstrates an insight solar PV Stand Alone system which is a practical model with a halogen light source. At different situations the performance of solar PV cells are analyzed. The system produces power with depending on the change in halogen light intensity & temperature. A theoretical & experimental analysis of the PV cell can be achieved. In this paper the I-V & P-V characteristic of the solar photovoltaic cells with changes in temperature and isolation have been showed. With...

  10. Microorganism selection and performance in bioslurry reactors treating PAH-contaminated soil.

    Science.gov (United States)

    Cassidy, D P; Hudak, A J

    2002-09-01

    A continuous-flow reactor (CSTR) and a soil slurry-sequencing batch reactor (SS-SBR) were operated in 81 vessels for 200 days to treat a soil contaminated with polycyclic aromatic hydrocarbons (PAH). Filtered slurry samples were used to quantify bulk biosurfactant concentrations and PAH emulsification. Concentrations of Corynebacterium aquaticum, Flavobacterium mizutaii, Mycobacterium gastri, Pseudomonas aeruginosa, and Pseudomonas putida were determined using fatty acid methyl ester (FAME) analysis. The CSTR and SS-SBR selected microbial consortia with markedly different surfactant-producing and PAH-degrading abilities. Biosurfactant levels in the SS-SBR reached 4 times the critical micelle concentration (CMC) that resulted in considerable emulsification of PAH. In contrast, CSTR operation resulted in nomeasurable biosurfactant production. Total PAH removal efficiency was 93% in the SS-SBR, compared with only 66% in the CSTR, and stripping of PAH was 3 times less in the SS-SBR. Reversing the mode of operation on day 100 caused a complete reversal in microbial consortia and in reactor performance by day 140. These results show that bioslurry reactor operation can be manipulated to control overall reactor performance.

  11. Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

    Science.gov (United States)

    Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

    2016-12-01

    This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent. Copyright © 2015 Elsevier Inc. All rights reserved.

  12. New steady-state microbial community compositions and process performances in biogas reactors induced by temperature disturbances

    DEFF Research Database (Denmark)

    Luo, Gang; De Francisci, Davide; Kougias, Panagiotis

    2015-01-01

    that stochastic factors had a minor role in shaping the profile of the microbial community composition and activity in biogas reactors. On the contrary, temperature disturbance was found to play an important role in the microbial community composition as well as process performance for biogas reactors. Although...... three different temperature disturbances were applied to each biogas reactor, the increased methane yields (around 10% higher) and decreased volatile fatty acids (VFAs) concentrations at steady state were found in all three reactors after the temperature disturbances. After the temperature disturbance...... in shaping the profile of the microbial community composition and activity in biogas reactors. New steady-state microbial community profiles and reactor performances were observed in all the biogas reactors after the temperature disturbance....

  13. Characteristics of biohydrogen production by ethanoligenens R{sub 3} isolated from continuous stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, A.Y.; Liu, K. [Northeast Forestry Univ., Harbin (China). School of Forestry; Li, Y.F. [Northeast Forestry Univ., Harbin (China). School of Forestry; Shanghai Univ. of Engineering Science (China). College of Chemistry and Chemical Engineering; Liu, B. [Northeast Forestry Univ., Harbin (China). School of Material Science and Engineering; Xu, J.L. [Shanghai Univ. of Engineering Science (China). College of Chemistry and Chemical Engineering

    2010-07-01

    This study investigated the fermentative hydrogen production characteristics of ethanoligenens R{sub 3} isolated from anaerobic sludge in a continuous stirred tank reactor. The effects of the initial pH value, the proportion of carbon and nitrogen sources, and the effects of fermentation temperature were investigated in a series of batch experiments. Substrates for the hydrogen production of glucose and peptone were used as carbon and nitrogen sources. Results of the experiments showed that a maximum hydrogen production yield of 834 mlH{sub 2}/L culture was obtained with a fermentation temperature of 35 degrees C and an initial pH value of 5.5. The maximum average hydrogen production rate of 10.87 mmolH{sub 2}/g cell dry weight per hour was obtained at a carbon-nitrogen source ratio of 3.3. The degradation efficiency of the glucose used as a carbon source ranged from 91.5 to 95.43 per cent during the conversion of glucose to hydrogen by the bacteria.

  14. Control characteristics of cryogenic distillation column with a feedback stream for fusion reactor

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Okuno, Kenji

    1997-01-01

    The control characteristics of the cryogenic distillation column with a feedback stream have been discussed based on computer simulation results. This column plays an important role in fusion reactor. A new control system was proposed from the simulation results. The flow rate of top product is determined from the composition and flow rate of a main feed stream by a feedforward control loop. The flow rates of the feedback stream and vapor stream within the column are proportionally changed with a corresponding change of feed flow rate. The flow rate of vapor stream within the column is further adjusted to maintain product purity by a feedback control loop. The proposed system can control the product purity for a large fluctuation of feed composition, a change of feed flow rate, and an increase or decrease of the number of total theoretical stages of the column. The control system should be designed for each column by considering its operating conditions and function. The present study gives us a basic procedure for the design method of the control system of the cryogenic distillation column. (author)

  15. Characteristics of dechlorination for LiCl salt by the surface temperature-controlled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, In Hak [Chungnam National University, Daejeon (Korea, Republic of); Park, Hwan Seo; Ahn, Soo Na; Eun, Hee Chul; Kim, In Tae; Cho, Yong Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    Molten salt waste is generated from a pyrochemical process to separate reusable U and TRU elements from a spent nuclear fuel. The spent lithium chloride waste is highly soluble in water and contains volatile radioactive elements such as Cs. However, these wastes are difficult to directly immobilize into durable matrix such as glass or ceramic wasteform for final disposal. ANL(Argonne National Laboratory) suggested the conversion of metal chloride into a sodalite for the immobilization of a chloride waste, glass-bonded sodalite, which was fabricated at about 915 .deg. C after mixing the salt-loaded zeolite and borosilicate glass powder. Although this wasteform shows high leach-resistance, the waste volume greatly increases. The previous study was to treat metal chloride wastes by using SAP(SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) materials. By using this method, the final waste volume reduced and leach-resistance was good. In this study, characteristics of dechlorination reaction of LiCl with an inorganic composite, SAP, was investigated by using a specific surface temperature-controlled reactor

  16. Characteristic Studies of Micron Zinc Particle Hydrolysis in a Fixed Bed Reactor

    Directory of Open Access Journals (Sweden)

    Lv Ming

    2015-09-01

    Full Text Available Zinc fuel is considered as a kind of promising energy sources for marine propeller. As one of the key steps for zinc marine energy power system, zinc hydrolysis process had been studied experimentally in a fixed bed reactor. In this study, we focus on the characteristics of micron zinc particle hydrolysis. The experimental results suggested that the steam inner diffusion is the controlling step of accumulative zinc particles hydrolysis reaction at a relative lower temperature and a relative higher water partial pressure. In other conditions, the chemical reaction kinetics was the controlling step. And two kinds of chemical reaction kinetics appeared in experiments: the surface reaction and the gas-gas reaction. The latter one occurs usually for larger zinc particles and high reaction temperature. Temperature seems to be one of the most important parameters for the dividing of different reaction mechanisms. Several parameters of the hydrolysis process including heating rate, water partial pressure, the particle size and temperature were also studied in this paper. Results show that the initial reaction temperature of zinc hydrolysis in fixed bed is about 410°C. And the initial reaction temperature increases as the heating rate increases and as the water partial pressure decreases. The total hydrogen yield increases as the heating rate decreases, as the water partial pressure increases, as the zinc particle size decreases, and as the reaction temperature increases. A hydrogen yield of more than 81.5% was obtained in the fixed bed experiments.

  17. Reactivity limitations on the performance of hybrid reactors

    International Nuclear Information System (INIS)

    Piera, M.; Martinez-Val, J.M.

    1994-01-01

    A neutronic theory for characterizing the hybrid blanket physics is used to show that hybrid performances are limited because of reactivity restrictions. The hybrid must always remain subcritical, even in abnormal conditions. For hybrids devoted to energy production by multiplication of the neutron source power, the restriction is particularly strong, and the electricity recirculation fraction to feed the plant can be too large for its economic feasibility. In hybrids used to breed fissile fuel, the power of LWR maintained by power unit of the hybrid is also limited (to a factor of 10, approx.). (author)

  18. The Jules Horowitz reactor, a new high performance European material testing reactor open to international users: present status and objectives

    International Nuclear Information System (INIS)

    Iracane, D.; Bignan, G.

    2010-01-01

    The development of nuclear power as a sustainable and competitive energy source will continue to require research and development of fuel and material behaviour under irradiation. This necessitates a high performance material testing reactor (MTR). Facing the obsolescence of most of the existing MTR in Europe, France decided a few years ago the construction of the RJH (Jules Horowitz reactor). RJH is designed, built and will be operated as an international user facility. A first set of experimental hosting devices is being designed. For instance, there are the in-core CALIPSO Nak integrated loop for material studies and other loops for fuel studies under nominal or off-normal or accidental conditions. The RJH international program will focus on the following subjects: -) fuel reliability, assessed through power ramps tests and post-irradiation examination; -) Loss of coolant tests done out-of-pile in a first phase and in-pile in a possible second phase; and -) source term tests addressing fission products release. The paper reports also the point of view of VATTENFALL (a Swedish power utility), as a potential European RJH user. (A.C.)

  19. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  20. Effect of inlet conditions on the performance of a palladium membrane reactor

    International Nuclear Information System (INIS)

    Birdsell, S.A.; Willms, R.S.; Arzu, P.; Costello, A.

    1997-10-01

    Palladium membrane reactors (PMR) will be used to remove tritium and other hydrogen isotopes from impurities, such as tritiated methane and tritiated water, in the exhaust of the International Thermonuclear Experimental Reactor. In addition to fusion-fuel processing, the PMR system can be used to recover tritium from tritiated waste water. This paper investigates the effect of inlet conditions on the performance of a PMR. A set of experiments were run to determine, independently, the effect of inlet compositions and residence time on performance. Also, the experiments were designed to determine if the injected form of hydrogen (CH 4 or H 2 O) effects performance. Results show that the PMR operates at optimal hydrogen recovery with a broad range of inlet compositions and performance is shown to increase with increased residence time. PMR performance is shown to be independent of whether hydrogen is injected in the form of CH 4 or H 2 O

  1. The operation characteristics of biohydrogen production in continuous stirred tank reactor with molasses

    Energy Technology Data Exchange (ETDEWEB)

    Hong, C.; Wei, H.; Jie-xuan, D.; Xin, Y.; Chuan-ping, Y. [Northeast Forestry Univ., Harbin (China). School of Forestry; Li, Y.F. [Northeast Forestry Univ., Harbin (China). School of Forestry; Shanghai Univ. Engineering, Shanghai (China). College of Chemistry and Chemical Engineering

    2010-07-01

    The anaerobic fermentation biohydrogen production in a continuous stirred tank reactor (CSTR) was investigated as a means for treating molasses wastewater. The research demonstrated that the reactor has the capacity of continuously producing hydrogen in an initial biomass (as volatile suspension solids) of 17.74 g/L, temperature of approximately 35 degrees Celsius, hydraulic retention time of 6 hours. The reactor could begin the ethanol-type fermentation in 12 days and realize stable hydrogen production. The study also showed that the CSTR reactor has a favourable stability even with an organic shock loading. The hydrogen yield and chemical oxygen demand (COD) increased, as did the hydrogen content.

  2. Cavity temperature and flow characteristics in a gas-core test reactor

    Science.gov (United States)

    Putre, H. A.

    1973-01-01

    A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

  3. Operational and safety characteristics of reactors with materials having remarkable indeterminateness in data

    International Nuclear Information System (INIS)

    Lelek, V.; Szatmary, Z.

    1999-01-01

    High Pu isotopes and minor actinides occur in contemporary reactors only in the very small amount and that is why we have not needed their data with high precise and it was also practically excluded to test them on the standard reactors measurements. On the contrary in the trans mutational technologies reactors consist of only such fissionable materials. Taking into account how hard was in the past to have good uranium libraries we can hardly rely that there will be such in our disposal before the start up the first experimental reactor for transmutation. (Authors)

  4. Assessing the influence of reactor system design criteria on the performance of model colon fermentation units.

    Science.gov (United States)

    Moorthy, Arun S; Eberl, Hermann J

    2014-04-01

    Fermentation reactor systems are a key platform in studying intestinal microflora, specifically with respect to questions surrounding the effects of diet. In this study, we develop computational representations of colon fermentation reactor systems as a way to assess the influence of three design elements (number of reactors, emptying mechanism, and inclusion of microbial immobilization) on three performance measures (total biomass density, biomass composition, and fibre digestion efficiency) using a fractional-factorial experimental design. It was determined that the choice of emptying mechanism showed no effect on any of the performance measures. Additionally, it was determined that none of the design criteria had any measurable effect on reactor performance with respect to biomass composition. It is recommended that model fermentation systems used in the experimenting of dietary effects on intestinal biomass composition be streamlined to only include necessary system design complexities, as the measured performance is not benefited by the addition of microbial immobilization mechanisms or semi-continuous emptying scheme. Additionally, the added complexities significantly increase computational time during simulation experiments. It was also noted that the same factorial experiment could be directly adapted using in vitro colon fermentation systems. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

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

    NARCIS (Netherlands)

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

    2010-01-01

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

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

    NARCIS (Netherlands)

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

    2010-01-01

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

  7. Evaluation of the dual digestion system 2: operation and performance of the pure oxygen aerobic reactor

    CSIR Research Space (South Africa)

    Messenger, JR

    1993-07-01

    Full Text Available In a comprehensive study of the performance of a full-scale (45 m3) pure oxygen autothermal thermophilic aerobic reactor of a sewage sludge dual digestion system, it was found that: Biological heat generation rate was directly proportional...

  8. A WIMS E analysis of zero energy experiments performed on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lancefield, M. J.; Broadhouse, B.; Woloch, F.

    1974-10-15

    UKAEA methods embodied in the WINS-E modular scheme of codes are described in their application to the analysis of zero energy experiments performed on the DRAGON reactor. Measured reactivity and reaction rate distributions are compared with the predictions of the analysis.

  9. The characteristics of extracellular polymeric substances and soluble microbial products in moving bed biofilm reactor-membrane bioreactor.

    Science.gov (United States)

    Duan, Liang; Jiang, Wei; Song, Yonghui; Xia, Siqing; Hermanowicz, Slawomir W

    2013-11-01

    The characteristics of extracellular polymeric substances (EPS) and soluble microbial products (SMP) in conventional membrane bioreactor (MBR) and in moving bed biofilm reactor-membrane bioreactors (MBBR-MBR) were investigated in long-term (170 days) experiments. The results showed that all reactors had high removal efficiency of ammonium and COD, despite very different fouling conditions. The MBBR-MBR with media fill ratio of 26.7% had much lower total membrane resistance and no obvious fouling were detected during the whole operation. In contrast, MBR and MBBR-MBR with lower and higher media fill experienced more significant fouling. Low fouling at optimum fill ratio may be due to the higher percentage of small molecular size (100 kDa) of EPS and SMP in the reactor. The composition of EPS and SMP affected fouling due to different O-H bonds in hydroxyl functional groups, and less polysaccharides and lipids. Copyright © 2013 Elsevier Ltd. All rights reserved.

  10. Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2016-11-15

    Highlights: • Simplified models with inclusion of the clad temperature are considered. • Boiling nonlinearity and core power removal have been modeled. • Method of multiple time scales has been used for nonlinear analysis to get the nature and amplitude of oscillations. • Incorporation of modeling complexities enhances the stability of system. • We find that reactors with higher nominal power are more desirable from the point of view of global stability. - Abstract: We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all

  11. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

  12. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

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

    1990-12-01

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

  13. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  14. Study on human-factors-engineering properties of reactor maintenance workers with protection suits, (2). Basic research on various biological characteristics in reactor maintenance simulation tests

    Energy Technology Data Exchange (ETDEWEB)

    Yoshino, K; Ishii, K; Nakasa, H [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Shigeta, S

    1980-11-01

    To ensure the safety of reactor maintenance workers and to reduce the radiation exposure through the enhancement of labor efficiency, it is needed to evaluate quantitatively work-stress levels of workers with radiation-protection suits. This paper presents the results of reactor-maintenance simulation tests in which the relationship between the work stress and biological characteristics is investigated for 5 pinds of model works done by testees without protection suits in an artificial climate chamber. Major results obtained are: (1) the selected model works are mostly evaluated to be relatively heavy through the measurement of RMR (Relative Metabolic Rate). (2) biological characteristics such as heart rate and respiratory volume under the model works have close relationship to RMR which is the cumulative quantity in relatively long time, and then they may become the real-time indicator for the work stress level. (3) such biological characteristics are greatly affected by the high-temperature work-environment which is often seen in workers with protection suits.

  15. Performance characteristics of solar air heater with surface mounted obstacles

    International Nuclear Information System (INIS)

    Bekele, Adisu; Mishra, Manish; Dutta, Sushanta

    2014-01-01

    Highlights: • Solar air heater with delta shaped obstacles have been studied. • Obstacle angle of incidence strongly affects the thermo-hydraulic performance. • Thermal performance of obstacle mounted collectors is superior to smooth collectors. • Thermo-hydraulic performance of the present SAH is higher than those in previous studies. - Abstract: The performance of conventional solar air heaters (SAHs) can be improved by providing obstacles on the heated wall (i.e. on the absorber plate). Experiments have been performed to collect heat transfer and flow-friction data from an air heater duct with delta-shaped obstacles mounted on the absorber surface and having an aspect ratio 6:1 resembling the conditions close to the solar air heaters. This study encompassed for the range of Reynolds number (Re) from 2100 to 30,000, relative obstacle height (e/H) from 0.25 to 0.75, relative obstacle longitudinal pitch (P l /e) from 3/2 to 11/2, relative obstacle transverse pitch (P t /b) from 1 to 7/3 and the angle of incidence (α) varied from 30° to 90°. The thermo-hydraulic performance characteristics of SAH have been compared with the previous published works and the optimum range of the geometries have been explored for the better performance of such air-heaters compared to the other designs of solar air heaters

  16. Model Of Emergency Department Nurse Performance Improvement Based on Association of Individual Characteristic, Organization Characteristic and Job Characteristic

    OpenAIRE

    Bogar, Maria Margaretha; Nursalam, Nursalam; Dewi, Yulis Setiya

    2017-01-01

    Introduction: Nursing care is integral part of health care and having important role in management of patient with emergency condition. The purpose of this research was to develop nurse performance improvement model based on individual, organization and job characteristics association in Emergency Department of RSUD dr TC Hillers Maumere. Method: This was an explanative survey by cross sectional approach held on July -August 2012. Respondents in this study were 22 nurses and 44 patients were ...

  17. Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

    2000-04-01

    The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

  18. Trickle bed reactor model to simulate the performance of commercial diesel hydrotreating unit

    Energy Technology Data Exchange (ETDEWEB)

    C. Murali; R.K. Voolapalli; N. Ravichander; D.T. Gokak; N.V. Choudary [Bharat Petroleum Corporation Ltd., Udyog Kendra (India). Corporate R& amp; D Centre

    2007-05-15

    A two phase mathematical model was developed to simulate the performance of bench scale and commercial hydrotreating reactors. Major hydrotreating reactions, namely, hydrodesulphurization, hydrodearomatization and olefins saturation were modeled. Experiments were carried out in a fixed bed reactor to study the effect of different process variables and these results were used for estimating kinetic parameters. Significant amount of feed vaporization (20-50%) was estimated under normal operating conditions of DHDS suggesting the importance of considering feed vaporization in DHDS modeling. The model was validated with plant operating data, under close to ultra low sulphur levels by correctly accounting for feed vaporization in heat balance relations and appropriate use of hydrodynamic correlations. The model could predict the product quality, reactor bed temperature profiles and chemical hydrogen consumption in commercial plant adequately. 14 refs., 7 figs., 6 tabs.

  19. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

    International Nuclear Information System (INIS)

    Tatone, O.S.; Meindl, P.; Taylor, G.F.

    1986-06-01

    A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

  20. Analytical model for performance verification of liquid poison injection system of a nuclear reactor

    International Nuclear Information System (INIS)

    Kansal, Anuj Kumar; Maheshwari, Naresh Kumar; Vijayan, Pallippattu Krishnan

    2014-01-01

    Highlights: • One-dimensional modelling of shut down system-2. • Semi-empirical correlation poison jet progression. • Validation of code. - Abstract: Shut down system-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1D) hydraulic code, COPJET is developed, to predict the performance of system by predicting progression of poison jet with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for advanced vertical pressure type reactor