WorldWideScience

Sample records for stable reactor behavior

  1. Policy on stable iodine prophylaxis following nuclear reactor accidents (1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This policy considers the alleviation of possible hazards that may arise from any radioiodines inhaled from a plume of fission products emanating from a nuclear reactor accident. Such a nuclear reactor may be land or ship-based. In any accident that releases radioiodines to the environment, one countermeasure that may need to be considered to reduce the effect of inhalation of radioiodines by persons downwind of the point of release is to provide those persons with tablets containing stable iodine. Both potassium iodide (KI) and potassium iodate (KIO 3 ) are recommended as effective prophylactics tablets for this purpose in Australia. Action levels, doses and contraindicatories are briefly outlined

  2. Application of stable adaptive schemes to nuclear reactor systems, (1)

    International Nuclear Information System (INIS)

    Fukuda, Toshio

    1978-01-01

    Parameter identification and adaptive control schemes are presented for a point reactor with internal feedbacks which lead to the nonlinearity of the overall system. Both are shown stable with new representation of the system, which corresponds to the nonminimal system representation, in the vein of the Model Reference Adaptive System (MRAS) via the Lyapunov's method. For the sake of the parameter identification, model parameters can be adjusted adaptively as soon as measurements start, while plant parameters can also adaptively be compensated through control input to reduce the output error between the model and the plant for the case of the adaptive control. In the case of the adaptive control, control schemes are presented for two cases, the case of the unknown decay constant of the delayed neutron and the case of the known constant. The adaptive control scheme for the latter case is shown extremely simpler than that for the former. Furthermore, when plant parameters vary slowly with time, computer simulations show that the proposed adaptive control scheme works satisfactorily enough to stabilize an unstable reactor and that it does even in the noise with small variance. (auth.)

  3. Behavior of water reactor fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1990-08-01

    This paper reviewed the fuels used widely in forms of (1) Zircaloy-sheathed UO 2 fuel in light water-commercial power reactor, (2) Zircaloy-sheathed PuO 2 -UO 2 fuel in plutonium-thermal reactor and advanced reactor (ATR), (3) aluminide and silicide fuel in Material Testing Reactors. From fundamental view points, physical/chemical properties and irradiation behaviors of both fuels and zircaloy claddings are briefly reviewed in chapters 1 and 2. Change of the fuel rod physical parameters with progress of burn-up are summed up in chapter 3. Some fuel troubles and failures encountered in past usage of worldwide LWR fuels are introduced with counterplans taken. In the last session of this chapter, recent results of R and D works have been carried out by fuel vendors are reviewed. Especially, in-core behaviors of PCI-remedy fuels developed to use for high burn-up extension and for load-follow operation are highlighted. Reactor accidents occurred through past forty years are surveyed and reviewed. Fuel behaviors during the reactivity initiated accident (RIA), the power-coolant mismatch (PCM), and the loss-of-coolant accident (LOCA) are taken into this review by using disclosed literatures. Safety criteria being used in Japanese licensing authorities are introduced relating to the fuel design limit. (author)

  4. Models of iodine behavior in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents.

  5. Copper stable isotopes to trace copper behavior in wetland systems.

    Science.gov (United States)

    Babcsányi, Izabella; Imfeld, Gwenaël; Granet, Mathieu; Chabaux, François

    2014-05-20

    Wetlands are reactive zones of the landscape that can sequester metals released by industrial and agricultural activities. Copper (Cu) stable isotope ratios (δ(65)Cu) have recently been used as tracers of transport and transformation processes in polluted environments. Here, we used Cu stable isotopes to trace the behavior of Cu in a stormwater wetland receiving runoff from a vineyard catchment (Alsace, France). The Cu loads and stable isotope ratios were determined in the dissolved phase, suspended particulate matter (SPM), wetland sediments, and vegetation. The wetland retained >68% of the dissolved Cu and >92% of the SPM-bound Cu, which represented 84.4% of the total Cu in the runoff. The dissolved Cu became depleted in (65)Cu when passing through the wetland (Δ(65)Cuinlet-outlet from 0.03‰ to 0.77‰), which reflects Cu adsorption to aluminum minerals and organic matter. The δ(65)Cu values varied little in the wetland sediments (0.04 ± 0.10‰), which stored >96% of the total Cu mass within the wetland. During high-flow conditions, the Cu flowing out of the wetland became isotopically lighter, indicating the mobilization of reduced Cu(I) species from the sediments and Cu reduction within the sediments. Our results demonstrate that the Cu stable isotope ratios may help trace Cu behavior in redox-dynamic environments such as wetlands.

  6. Creep behavior of materials for high-temperature reactor application

    International Nuclear Information System (INIS)

    Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

    1984-01-01

    Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

  7. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  8. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  9. Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

    1994-01-01

    Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

  10. Actinide behavior in the integral fast reactor

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

  11. Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

    International Nuclear Information System (INIS)

    Irianto, Djoko; Kanji, T.; Kukita, Y.

    1996-01-01

    An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

  12. Identification of the autotrophic denitrifying community in nitrate removal reactors by DNA-stable isotope probing.

    Science.gov (United States)

    Xing, Wei; Li, Jinlong; Cong, Yuan; Gao, Wei; Jia, Zhongjun; Li, Desheng

    2017-04-01

    Autotrophic denitrification has attracted increasing attention for wastewater with insufficient organic carbon sources. Nevertheless, in situ identification of autotrophic denitrifying communities in reactors remains challenging. Here, a process combining micro-electrolysis and autotrophic denitrification with high nitrate removal efficiency was presented. Two batch reactors were fed organic-free nitrate influent, with H 13 CO 3 - and H 12 CO 3 - as inorganic carbon sources. DNA-based stable-isotope probing (DNA-SIP) was used to obtain molecular evidence for autotrophic denitrifying communities. The results showed that the nirS gene was strongly labeled by H 13 CO 3 - , demonstrating that the inorganic carbon source was assimilated by autotrophic denitrifiers. High-throughput sequencing and clone library analysis identified Thiobacillus-like bacteria as the most dominant autotrophic denitrifiers. However, 88% of nirS genes cloned from the 13 C-labeled "heavy" DNA fraction showed low similarity with all culturable denitrifiers. These findings provided functional and taxonomical identification of autotrophic denitrifying communities, facilitating application of autotrophic denitrification process for wastewater treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

    Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

  14. Transient behavior of a nuclear reactor coupled to an accelerator

    Science.gov (United States)

    Sadineni, Suresh Babu

    Accelerator Driven Systems (ADS) present one of the most viable solutions for transmutation and effective utilization of nuclear fuel. Spent fuel from reactors will be partitioned to separate plutonium and other minor actinides to be transmuted in the ADS. Without the ADS, minor actinides must be stored at a geologic repository for long periods of time. One problem with ADS is understanding the control issues that arise when coupling an accelerator to a reactor. "ADSTRANS" was developed to predict the transient behavior of a nuclear reactor coupled to an accelerator. It was based on MCNPX, a radiation transport code developed at the LANL, and upon a numerical model of the neutron transport equation. MCNPX was used to generate the neutron "source" term that occurs when the accelerator is fired. ADSTRANS coupled MCNPX to a separate finite difference code that solved the transient neutron transport equation. A cylindrical axisymmetric reactor with steel shielding was considered for this analysis. Multiple neutron energy groups, neutron precursor groups and neutron poisons were considered. ENDF/B cross-section data obtained through MCNPX was also employed. The reactor was assumed to be isothermal and near zero power level. Unique features of this code are: (1) it predicts the neutron behavior of an ADS for different reactor geometry, material concentration, both electron and proton particle accelerators, and target material, (2) it develops input files for MCNPX to simulate neutron production, runs MCNPX, and retrieves information from the MCNPX output files. Neutron production predicted by MCNPX for a 20 MeV electron accelerator and lead target was compared with experimental data from the Idaho Accelerator Center and found to be in good agreement. The spatial neutron flux distribution and transient neutron flux in the reactor as predicted by the code were compared with analytical solutions and found to be in good agreement. Fuel burnup and poison buildup were also as

  15. Tritium behavior in the Caisson, a simulated fusion reactor room

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Yamada, Masayuki; Suzuki, Takumi; O'hira, Shigeru; Nakamura, Hirofumi; Shu, Weimin; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Konishi, Satoshi; Nishi, Masataka

    2000-01-01

    In order to confirm tritium confinement ability in the deuterium-tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called 'Caisson', at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak-tight vessel of 12 m 3 , simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code

  16. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Moreira, F.J.

    1984-01-01

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

  17. Dynamic Behavior of Reverse Flow Reactor for Lean Methane Combustion

    Directory of Open Access Journals (Sweden)

    Yogi W. Budhi

    2014-09-01

    Full Text Available The stability of reactor operation for catalytic oxidation of lean CH4 has been investigated through modeling and simulation, particularly the influence of switching time and heat extraction on reverse flow reactor (RFR performance. A mathematical model of the RFR was developed, based on one-dimensional pseudo-homogeneous model for mass and heat balances, incorporating heat loss through the reactor wall. The configuration of the RFR consisted of inert-catalyst-inert, with or without heat extraction that makes it possible to store the energy released by the exothermic reaction of CH4 oxidation. The objective of this study was to investigate the dynamic behavior of the RFR for lean methane oxidation and to find the optimum condition by exploring a stability analysis of the simple reactor. The optimum criteria were defined in terms of CH4 conversion, CH4 slip, and heat accumulation in the RFR. At a switching time of 100 s, the CH4 conversion reached the maximum value, while the CH4 slip attained its minimum value. The RFR could operate autothermally with positive heat accumulation, i.e. 0.02 J/s. The stability of the RFR in terms of heat accumulation was achieved at a switching time of 100 s.

  18. Transmutation of technetium into stable ruthenium in high flux conceptual research reactor

    International Nuclear Information System (INIS)

    Amrani, N.; Boucenna, A.

    2007-01-01

    The effectiveness of transmutation for the long lived fission product technetium-99 in high flux research reactor, considering its large capture cross section in thermal and epithermal region is evaluated. The calculation of Ruthenium concentration evolution under irradiation was performed using Chain Solver 2.20 code. The approximation used for the transmutation calculation is the assumption that the influence of change in irradiated materials structures on the reactor operator mode characteristics is insignificant. The results on Technetium transmutation in high flux research reactor suggested an effective use of this kind of research reactors. The evaluation brings a new concept of multi-recycle Technetium transmutation using HFR T RAN (High Flux Research Reactor for Transmutation)

  19. Use of stable emulsion to improve stability, activity, and enantioselectivity of lipase immobilized in a membrane reactor.

    Science.gov (United States)

    Giorno, L; Li, N; Drioli, E

    2003-12-20

    The enantiocatalytic performance of immobilized lipase in an emulsion membrane reactor using stable emulsion prepared by membrane emulsification technology was studied. The production of optical pure (S)-naproxen from racemic naproxen methyl ester was used as a model reaction system. The O/W emulsion, containing the substrate in the organic phase, was fed to the enzyme membrane reactor from shell-to-lumen. The enzyme was immobilized in the sponge layer (shell side) of capillary polyamide membrane with 50 kDa cut-off. The aqueous phase was able to permeate through the membrane while the microemulsion was retained by the thin selective layer. Therefore, the substrate was kept in the enzyme-loaded membrane while the water-soluble product was continuously removed from the reaction site. The results show that lipase maintained stable activity during the entire operation time (more than 250 h), showing an enantiomeric excess (96 +/- 2%) comparable to the free enzyme (98 +/- 1%) and much higher compared to similar lipase-loaded membrane reactors used in two-separate phase systems (90%). The results demonstrate that immobilized enzymes can achieve high stability as well as high catalytic activity and enantioselectivity. Copyright 2003 Wiley Periodicals, Inc.

  20. A neural network to predict reactor core behaviors

    International Nuclear Information System (INIS)

    Juan Jose Ortiz-Servin; Jose Alejandro Castillo; Pelta, David A.

    2014-01-01

    The global fuel management problem in BWRs (Boiling Water Reactors) can be understood as a very complex optimization problem, where the variables represent design decisions and the quality assessment of each solution is done through a complex and computational expensive simulation. This last aspect is the major impediment to perform an extensive exploration of the design space, mainly due to the time lost evaluating non promising solutions. In this work, we show how we can train a Multi-Layer Perceptron (MLP) to predict the reactor behavior for a given configuration. The trained MLP is able to evaluate the configurations immediately, thus allowing performing an exhaustive evaluation of the possible configurations derived from a stock of fuel lattices, fuel reload patterns and control rods patterns. For our particular problem, the number of configurations is approximately 7.7 x 10 10 ; the evaluation with the core simulator would need above 200 years, while only 100 hours were required with our approach to discern between bad and good configurations. The later were then evaluated by the simulator and we confirm the MLP usefulness. The good core configurations reached the energy requirements, satisfied the safety parameter constrains and they could reduce uranium enrichment costs. (authors)

  1. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  2. Switching behavior and novel stable states of magnetic hexagonal nanorings

    Energy Technology Data Exchange (ETDEWEB)

    Yasir Rafique, M., E-mail: myasir.rafique@ciitlahore.edu.pk [Department of Physics, COMSATS Institute of Information Technology, Lahore 54000 (Pakistan); Pan, Liqing; Guo, Zhengang [College of Science and Research Institute for New Energy, China Three Gorges University, Yichang 443002 (China)

    2017-06-15

    Micromagnetic simulations for Cobalt hexagonal shape nanorings show onion (O) and vortex state (V) along with new state named “tri-domain state”. The tri-domain state is observed in sufficiently large width of ring. The magnetic reversible mechanism and transition of states are explained with help of vector field display. The transitions from one state to other occur by propagation of domain wall. The vertical parts of hexagonal rings play important role in developing the new “tri-domain” state. The behaviors of switching fields from onion to tri-domain (HO-Tr), tri-domain to vortex state (HTr-V) and vortex to onion state and “states size” are discussed in term of geometrical parameter of ring.

  3. Rheology behaviors of stable electrohydrodynamic direct-write jet

    Science.gov (United States)

    Wang, Xiang; Zheng, Gaofeng; Xu, Lei; Wang, Han; Li, Wenwang

    2016-10-01

    Electrohydrodynamic direct-write (EDW) is a novel direct-write technology to fabricate micro/nano-structures from viscoelastic solution, which had displayed great application potential in organic electronic device. Due to the shorter spinneret to substrate distance, the rheology behaviors of EDW charged jet played an important role in defining the line width or diameter of the direct-written micro/nano-structures. High speed camera is utilized to observe the rheology process of EDW charged jet, and solidified jets are measured by SEM that offers a quantitative method to investigate the diameter evolution of jet. The diameter of charged jet and nanofiber injected from solid probe increase with the increasing of polymer solution concentration. Attribute to the larger diameter and higher solvent content, charged jet injected from hollow nozzle displayed greater fluid viscoelasticity, and then stretched into micro structure of flat film under the gravitation on the substrate. The diameter of charged jet and line width of thin film injected from nozzle decrease with the increasing of polymer concentration.

  4. Behavior and Convergence of Wasserstein Metric in the Framework of Stable Distributions

    Czech Academy of Sciences Publication Activity Database

    Omelchenko, Vadym

    2012-01-01

    Roč. 2012, č. 30 (2012), s. 124-138 ISSN 1212-074X R&D Projects: GA ČR GAP402/10/0956 Institutional research plan: CEZ:AV0Z10750506 Institutional support: RVO:67985556 Keywords : Wasserstein Metric * Stable Distributions * Empirical Distribution Function Subject RIV: BB - Applied Statistics, Operational Research http://library.utia.cas.cz/separaty/2013/E/omelchenko-behavior and convergence of wasserstein metric in the framework of stable distributions.pdf

  5. Prospective Analysis of Behavioral Economic Predictors of Stable Moderation Drinking Among Problem Drinkers Attempting Natural Recovery.

    Science.gov (United States)

    Tucker, Jalie A; Cheong, JeeWon; Chandler, Susan D; Lambert, Brice H; Pietrzak, Brittney; Kwok, Heather; Davies, Susan L

    2016-12-01

    As interventions have expanded beyond clinical treatment to include brief interventions for persons with less severe alcohol problems, predicting who can achieve stable moderation drinking has gained importance. Recent behavioral economic (BE) research on natural recovery has shown that active problem drinkers who allocate their monetary expenditures on alcohol and saving for the future over longer time horizons tend to have better subsequent recovery outcomes, including maintenance of stable moderation drinking. This study compared the predictive utility of this money-based "Alcohol-Savings Discretionary Expenditure" (ASDE) index with multiple BE analogue measures of behavioral impulsivity and self-control, which have seldom been investigated together, to predict outcomes of natural recovery attempts. Community-dwelling problem drinkers, enrolled shortly after stopping abusive drinking without treatment, were followed prospectively for up to a year (N = 175 [75.4% male], M age = 50.65 years). They completed baseline assessments of preresolution drinking practices and problems, analogue behavioral choice tasks (Delay Discounting, Melioration-Maximization, and Alcohol Purchase Tasks), and a Timeline Followback interview including expenditures on alcohol compared to voluntary savings (ASDE index) during the preresolution year. Multinomial logistic regression models showed that, among the BE measures, only the ASDE index predicted stable moderation drinking compared to stable abstinence or unstable resolutions involving relapse. As hypothesized, stable moderation was associated with more balanced preresolution allocations to drinking and savings (odds ratio = 1.77, 95% confidence interval = 1.02 to 3.08, p behavior regulation processes than abstinence. The ASDE's unique predictive utility may rest on its comprehensive representation of contextual elements to support this patterning of behavioral allocation. Stable low-risk drinking, but not abstinence

  6. Stable Early Maternal Report of Behavioral Inhibition Predicts Lifetime Social Anxiety Disorder in Adolescence

    Science.gov (United States)

    Chronis-Tuscano, Andrea; Degnan, Kathryn Amey; Pine, Daniel S.; Perez-Edgar, Koraly; Henderson, Heather A.; Diaz, Yamalis; Raggi, Veronica L.; Fox, Nathan A.

    2009-01-01

    The odds of a lifetime diagnosis of social anxiety disorder increased by 3.79 times for children who had a stable report of behavioral inhibition from their mothers. This finding has important implications for the early identification and prevention of social anxiety disorder.

  7. A behavioral and cognitive profile of clinically stable HIV-infected children

    NARCIS (Netherlands)

    Nozyce, ML; Lee, SS; Wiznia, A; Nachman, S; Mofenson, LM; Smith, ME; Yogev, R; McIntosh, K; Stanley, K; Pelton, S

    OBJECTIVE. The purpose of this research was to characterize behavioral and cognitive profiles of clinically and immunologically stable antiretroviral-experienced HIV-infected children. METHODS. Two hundred seventy-four previously treated HIV-infected children aged 2 to 17 years were assessed for

  8. Comparative biogeochemical behaviors of iron-55 and stable iron in the marine environment

    International Nuclear Information System (INIS)

    Weimer, W.C.; Langford, J.C.; Jenkins, C.E.

    1978-01-01

    Studies of atmospheric aerosols have demonstrated that much of the 55 Fe associated with the aerosol input to the oceans is present as either an amorphous or hydrous iron oxide or as very small particulate species attached to the surfaces of the large aerosol particles. By comparison, nearly all of the stable iron is bound in the mineral phase of aerosol particles. This difference in the chemical and physical forms of the radioactive and stable iron isotopes results in the 55 Fe being more biologically available than is the stable iron. This difference in availability is responsible for the transfer of a much higher specific activity 55 Fe to certain ocean organisms and man relative to the specific activity of the total aerosol or of sea water. This differential biological uptake of the radioactive element and its stable element counterpart points out that natural levels of stable elements in the marine environment may not effectively dilute radioelements or other stable elements of anthropogenic sources. The effectiveness of dilution by natural sources depends on the chemical and physical forms of the materials in both the source terms and the receiving environments. The large difference in specific activities of 55 Fe in aerosols and sea water relative to ocean organisms reflects the independent behaviors of 55 Fe and stable iron

  9. Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

    International Nuclear Information System (INIS)

    Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

    2009-01-01

    The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

  10. Feedback control of a primary pump for safe and stable operation of a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Imai, S.; Masaoka, H.; Tamaki, M.; Kukita, Y.

    1993-01-01

    A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for the stable normal operation and a loss-of-feedwater test for the safe shutdown in an accident condition, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (orig.)

  11. Thermal-hydraulic experiment for safe and stable operation of a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Imai, S.; Masaoka, H.; Irianto, I.D.; Kohketsu, H.; Tamaki, M.; Anoda, Y.; Murata, H.; Kukita, Y.

    1992-01-01

    A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for normal operation simulation and a loss of feedwater test for an accident condition simulation, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (author)

  12. Behavior of low-burnup metallic fuels for the integral fast reactor at elevated temperatures in ex-reactor tests

    International Nuclear Information System (INIS)

    Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

    1991-07-01

    A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab

  13. 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.)

  14. 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.)

  15. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  16. Behavior of stainless steels in pressurized water reactor primary circuits

    International Nuclear Information System (INIS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-01-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  17. Behavior of 241Am in fast reactor systems - a safeguards perspective

    International Nuclear Information System (INIS)

    Beddingfield, David H.; Lafleur, Adrienne M.

    2009-01-01

    Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of 241 Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased (α,n) production in oxide fuels from the 241 Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of 241 Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of 241 Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of 241 Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

  18. Study of hydrodynamic behavior of a laboratory scale upflow anaerobic fixed film fixed bed reactor.

    Science.gov (United States)

    Tembhurkar, A R; Mhaisalkar, V A

    2006-04-01

    Application of anaerobic fixed film fixed bed reactor (AFFFB), containing mixed population of bacteria, helps in bacterial immobilization as biofilm on a support medium in the reactor overcoming the problem of biomass washout. Knowledge of hydrodynamics characteristics helps in better understanding and evaluating the performance of any reactor. This paper presents residence time distribution (RTD) studies carried out to investigate the hydrodynamic behavior of AFFFB reactor under the influence of varied hydraulic loading resulting in HRTs in range 24 hrs to 6 hrs. Results have demonstrated that the flow pattern is more close to a plug flow type and vary with varying intensity of dispersion and the extent of dispersion increases with decrease in HRTs.

  19. Calibration of hydrodynamic behavior and biokinetics for TOC removal modeling in biofilm reactors under different hydraulic conditions.

    Science.gov (United States)

    Zeng, Ming; Soric, Audrey; Roche, Nicolas

    2013-09-01

    In this study, total organic carbon (TOC) biodegradation was simulated by GPS-X software in biofilm reactors with carriers of plastic rings and glass beads under different hydraulic conditions. Hydrodynamic model by retention time distribution and biokinetic measurement by in-situ batch test served as two significant parts of model calibration. Experimental results showed that TOC removal efficiency was stable in both media due to the enough height of column, although the actual hydraulic volume changed during the variation of hydraulic condition. Simulated TOC removal efficiencies were close to experimental ones with low theil inequality coefficient values (below 0.15). Compared with glass beads, more TOC was removed in the filter with plastic rings due to the larger actual hydraulic volume and lower half saturation coefficient in spite of its lower maximum specific growth rate of biofilm, which highlighted the importance of calibrating hydrodynamic behavior and biokinetics. Copyright © 2013 Elsevier Ltd. All rights reserved.

  20. The close relation between Lactococcus and Methanosaeta is a keystone for stable methane production from molasses wastewater in a UASB reactor.

    Science.gov (United States)

    Kim, Tae Gwan; Yun, Jeonghee; Cho, Kyung-Suk

    2015-10-01

    The up-flow anaerobic sludge blanket (UASB) reactor is a promising method for the treatment of high-strength industrial wastewaters due to advantage of its high treatment capacity and settleable suspended biomass retention. Molasses wastewater as a sugar-rich waste is one of the most valuable raw material for bioenergy production due to its high organic strength and bioavailability. Interpretation for complex interactions of microbial community structures and operational parameters can help to establish stable biogas production. RNA-based approach for biogas production systems is recommended for analysis of functionally active community members which are significantly underestimated. In this study, methane production and active microbial community were characterized in an UASB reactor using molasses wastewater as feedstock. The UASB reactor achieved a stable process performance at an organic loading rate of 1.7~13.8-g chemical oxygen demand (COD,·L(-1) day(-1); 87-95 % COD removal efficiencies), and the maximum methane production rate was 4.01 L-CH4·at 13.8 g-COD L(-1) day(-1). Lactococcus and Methanosaeta were comprised up to 84 and 80 % of the active bacterial and archaeal communities, respectively. Network analysis of reactor performance and microbial community revealed that Lactococcus and Methanosaeta were network hub nodes and positively correlated each other. In addition, they were positively correlated with methane production and organic loading rate, and they shared the other microbial hub nodes as neighbors. The results indicate that the close association between Lactococcus and Methanosaeta is responsible for the stable production of methane in the UASB reactor using molasses wastewater.

  1. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  2. Tribological behavior of inconel 718 in sodium cooled reactor environments

    International Nuclear Information System (INIS)

    Wilson, W.L.; Galioto, T.A.; Schrock, S.L.

    1976-01-01

    Results of the present study on the tribological behavior of Inconel 718 in a sodium environment are summarized as follows: (a) Stroke lengths less than or equal to one-half the test pin diameter result in higher friction coefficients. (b) At elevated temperatures, the formation of a lubricative surface film can significantly influence the frictional behavior. (c) Tangential forces present during static dwell periods result in greater bonding tendencies. (d) Increasing contact pressure during static dwell periods results in lower breakaway friction coefficients

  3. Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

    1989-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

  4. Adsorptive control of water in esterification with immobilized enzymes: II. fixed-bed reactor behavior.

    Science.gov (United States)

    Mensah, P; Gainer, J L; Carta, G

    1998-11-20

    Experimental and theoretical studies are conducted to understand the dynamic behavior of a continuous-flow fixed-bed reactor in which an esterification is catalyzed by an immobilized enzyme in an organic solvent medium. The experimental system consists of a commercial immobilized lipase preparation known as Lipozyme as the biocatalyst, with propionic acid and isoamyl alcohol (dissolved in hexane) as the reaction substrates. A complex dynamic behavior is observed experimentally as a result of the simultaneous occurrence of reaction and adsorption phenomena. Both propionic acid and water are adsorbed by the biocatalyst resulting in lower reaction rates. In addition, an excessive accumulation of water in the reactor leads to a rapid irreversible inactivation of the enzyme. A model based on previously-obtained adsorption isotherms and kinetic expressions, as well as on adsorption rate measurements obtained in this work, is used to predict the concentration and thermodynamic activity of water along the reactor length. The model successfully predicts the dynamic behavior of the reactor and shows that a maximum thermodynamic activity of water occurs at a point at some distance from the reactor entrance. A cation exchange resin in sodium form, packed in the reactor as a selective water adsorbent together with the catalyst particles, is shown to be an effective means for preventing an excessive accumulation of water formed in the reaction. Its use results in longer cycle times and greater productivity. As predicted by the model, the experimental results show that the water adsorbed on the catalyst and on the ion exchange resin can be removed with isoamyl alcohol with no apparent loss in enzyme activity. Copyright 1998 John Wiley & Sons, Inc.

  5. Dynamic behavior of the HTR-10 reactor: Dual temperature feedback model

    Directory of Open Access Journals (Sweden)

    Hosseini Seyed Ali

    2015-01-01

    Full Text Available The current work aims at presenting a simple model for PBM-type reactors' dynamic behavior analysis. The proposed model is based on point kinetics equations coupled with feedbacks from fuel and moderator temperatures. The temperature reactivity coefficients were obtained through MCNP code and via available experimental data. Parameters such as heat capacity and heat conductivity were carefully analyzed and the final system of equations was numerically solved. The obtained results, while in partial agreement with previously proposed models, suggest lower sensitivity to step reactivity insertion as compared to other reactor designs and inherent safety of the design.

  6. A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tanaka, Toshiyuki

    1995-11-01

    A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

  7. Aerosol behavior in the reactor containment building during severe accident

    International Nuclear Information System (INIS)

    Berthion, Y.; Lhiaubet, G.; Gauvain, J.

    1984-07-01

    Thermohydraulic behavior inside a PWR containment during severe accident depends on decay heat transferred to the sump water by aerosol gravitational settling and deposition. Conversely, aerosol behavior depends on thermal hydraulic conditions, especially atmosphere moisture for soluble aerosol GsI, and CsOH. Therefore, a small iterative procedure between thermo-hydraulic and aerosol calculations has been performed in order to evaluate the importance of this coupling between the two phenomena. In this paper, it is shown that with this procedure and using our codes JERICHO, RICOCHET and AEROSOLS/B1, the steam condensation on aerosols is an important phenomenon for a correct estimation of the attenuation factor of the suspended mass of aerosols in the airborne of the containment. Then, we have a more realistic assessment of the source term released by the containment

  8. Performance and microbial population dynamics during stable operation and reactivation after extended idle conditions in an aerobic granular sequencing batch reactor.

    Science.gov (United States)

    He, Qiulai; Zhang, Wei; Zhang, Shilu; Zou, Zhuocheng; Wang, Hongyu

    2017-08-01

    The evolution of removal performance and bacterial population dynamics of an aerobic granular sequencing batch reactor were investigated during stable operation and reactivation after prolonged storage. The system was run for a period of 130days including the stable condition phase, storage period and the subsequent reactivation process. Excellent removal performance was obtained during the stable operation period, which was decayed by the extended idle conditions. The removal efficiencies for both carbon and nitrogen decayed while phosphorus removal remained unaffected. Both granules structure and physical properties could be fully restored. Microbial populations shifted sharply and the storage perturbations irreversibly altered the microbial communities at different levels. Extracellular polymeric substances (especially protein) and key groups were identified as contributors for storage and re-startup of the aerobic granular system. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Distribution and behavior of radionuclides and stable elements in Lake Obuchi

    International Nuclear Information System (INIS)

    Ueda, Shinji; Hasegawa, Hidenao; Hisamatsu, Shun'ichi; Inaba, Jiro

    2000-01-01

    Distribution and behavior of radionuclides and related stable elements in the lake water of brackish Lake Obuchi were investigated by field observations. Concentrations of 238 U and stable elements were measured at various points in the lake, and compiled to obtain the elemental distributions and variation characteristics. The concentrations of 238 U in the lake water were higher in areas nearer to the Pacific Ocean, and correlated well with those of Na, K, Ca, Mg and Sr (r = 0.86 to 0.92). These observations implied that 238 U in the lake originated from seawater. The bottom layer water was reductive during July and September (stratified period) in deep areas (> 3 m). In this condition, concentrations of PO 4 3- -P, NH 4 + -N, Fe and Mn in the water increased. Concentration ratios of 238 U to those of Na strongly suggested the following conclusions. The concentrations of 238 U in the turn-over period were represented by a simple mixture of seawater and fresh water. However, in the stratified period, part of the 238 U was lost from the seawater near the bottom of the lake due to the reductive condition. (author)

  10. Behavior of radionuclides and stable elements in the ecosystem in brackish Lake Obuchi

    International Nuclear Information System (INIS)

    Kondo, Kunio; Ueda, Shinji; Kawabata, Hitoshi; Hasegawa, Hidenao

    1998-01-01

    The behaviors and movements of radionuclides in a brackish environment, Lake Obuchi and its ecosystem were investigated with an aim to evaluate the safety of radioactive materials discharged into the lake. the organic materials in the sediment samples taken from Lake Obuchi were mainly composed of plant planktons and contained various radionuclides; 0.12-1.08 μg/g of 232 Th, 1.69-2.44 μg/g of 238 U, 0.0112-0.0176 Bq/g dry weight. The concentration of stable element was 145.6-357.3 μg/g for Mn, 82.5-128.2 μg/g for Sr, 0.81-1.65 μg/g for Cs and 11.2-16.5 μg/g dry weight for Pb. Plant planktons were found to mediate the accumulation of elements into lake sediments. Therefore, the species composition, distribution and ecological positioning of those planktons were important factors for the behaviors and movements of such elements. Then, the relationship between the concentrations of 137 Cs and Cs was studied with phytoplankton and a positive correlation (r=0.95) was found at both concentrations, suggesting that the adsorptions of the two elements onto the lake sediments were similar. (M.N.)

  11. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    International Nuclear Information System (INIS)

    Heams, T.J.; Williams, D.A.; Johns, N.A.; Mason, A.; Bixler, N.E.; Grimley, A.J.; Wheatley, C.J.; Dickson, L.W.; Osborn-Lee, I.; Domagala, P.; Zawadzki, S.; Rest, J.; Alexander, C.A.; Lee, R.Y.

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided

  12. Comparative study on two different seal surface structure for reactor pressure vessel sealing behavior

    International Nuclear Information System (INIS)

    Chen Jun; Xiong Guangming; Deng Xiaoyun

    2014-01-01

    The seal surface structure is very important to reactor pressure vessel (RPV) sealing behavior. In this paper, two 3-D RPV sealing analysis finite models have been established with different seal surface structures, in order to study the influence of two structures. The separation of RPV upper and lower flanges, bolt loads and etc. are obtained, which are used to evaluate the sealing behavior of the RPV. Meanwhile, the comparative analysis of safety margin of two seal surface structural had been done, which provides the theoretical basis for RPV seal structure design optimization. (authors)

  13. 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.)

  14. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is a neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch. 1 claim, 16 figures

  15. Experimental investigation on flow behavior during start-up of a heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin; Zhang Youjie

    1997-09-01

    An experimental simulation study on the transition from pressurized to boiling operation of a low-temperature, natural circulation nuclear heating reactor (5 MW) developed by INET of Tsinghua University is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instabilities, namely geyser instability, flashing instability and low-steam quality density wave instability on the transition from pressurized to boiling operation is described. The mechanism of flashing instability, which has never been studied well on this field, is especially interpreted. It is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: (1) increasing of initial pressure by means of a noncondensable gas (N 2 ), which is a very effective method to eliminate geyser instability and flashing instability at lower pressure. (2)start-up of the reactor at this pressurized condition with a constant heat flux under the limited value of q = 0.15 MW·m -2 , which controls the exit temperature of the heated section below the one of net vapor generation, the low steam quality density wave oscillation can be avoided. (3) transition to a lower pressure, boiling operation. The method of transition with low-heat flux and low-inlet subcooling is proposed: at pressurized operation condition, by reducing the heat flux to its lowest level, releasing the noncondensable gas and increasing the heat flux gradually (dq/dt -2 ·min -1 ), during which the low-steam quality density wave oscillation can be prevented from occurring, then the boiling operation condition can be achieved through adjusting the heat flux and inlet subcooling to their designed value. A stable transition from pressurized to boiling operation of the 5 MW reactor is achieved by careful selection of the thermohydraulic parameters. (7 refs., 7 figs., 1

  16. Sicral F1 graphite-core fuel element behavior in power reactors

    International Nuclear Information System (INIS)

    Rendu, M.

    1987-02-01

    Over 500 000 Sicral F1 graphite-core fuel elements have been manufactured by COGEMA to date and irradiated in GCR power reactors. Since 1963, this type of fuel element has been thoroughly investigated in design studies, in-core and out-of-core tests and post-mortem examinations. This report reviews the current state of knowledge on the irradiation behavior of the components under normal operating conditions and in incident situations (e.g. clad failure). It discusses how this work has led to optimization of the thermal, mechanical metallurgical and neutronic performance in order to obtain a can failure probability of less than 1.6 x 10 -5 , and defines general operating procedures for reactor implementation of this type of fuel element [fr

  17. Vibration monitoring of the mechanical behavior of the internal structures of PWR reactors

    International Nuclear Information System (INIS)

    Assedo, R.; Carre, J.C.; Sol, J.C.

    1979-01-01

    The internal structures of pressurized water reactors are the seat of vibrations induced by fluctuations in primary fluid flow. A knowledge of these phenomena is indispensable in order to ensure that the structures are in proper mechanical order. It can also be used for operational monitoring. This paper describes all the methods developed and the results already achieved in this domain. The first part deals with tests on mockup associated with the calculation models which afforded a good knowledge of the vibrational characteristics of the internal structures, as well as the measurements made during hot tests of certain reactors which made it possible to qualify these models on real structures. The second part describes the means of detection (neutron noise, external accelerometers) as well as the processing methods used in the follow-up. A few typical results obtained on site are then presented. Finally, the general principles of operational monitoring of the mechanical behavior of the internal structures are described [fr

  18. Vibration behavior of fuel-element vibration suppressors for the advanced power reactor

    Science.gov (United States)

    Adams, D. W.; Fiero, I. B.

    1973-01-01

    Preliminary shock and vibration tests were performed on vibration suppressors for the advanced power reactor for space application. These suppressors position the fuel pellets in a pin type fuel element. The test determined the effect of varying axial clearance on the behavior of the suppressors when subjected to shock and vibratory loading. The full-size suppressor was tested in a mockup model of fuel and clad which required scaling of test conditions. The test data were correlated with theoretical predictions for suppressor failure. Good agreement was obtained. The maximum difference with damping neglected was about 30 percent. Neglecting damping would result in a conservative design.

  19. Behavior and fate of copper ions in an anammox granular sludge reactor and strategies for remediation

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zheng-Zhe; Deng, Rui; Cheng, Ya-Fei; Zhou, Yu-Huang; Buayi, Xiemuguli [College of Life and Environmental Sciences, Hangzhou Normal University, Hangzhou 310036 (China); Key Laboratory of Hangzhou City for Ecosystem Protection and Restoration, Hangzhou Normal University, Hangzhou 310036 (China); Zhang, Xian; Wang, Hui-Zhong [College of Life and Environmental Sciences, Hangzhou Normal University, Hangzhou 310036 (China); Jin, Ren-Cun, E-mail: jrczju@aliyun.com [College of Life and Environmental Sciences, Hangzhou Normal University, Hangzhou 310036 (China); Key Laboratory of Hangzhou City for Ecosystem Protection and Restoration, Hangzhou Normal University, Hangzhou 310036 (China)

    2015-12-30

    Highlights: • The Cu partition in an anammox UASB reactor was predicted by models. • The distribution and form dynamics of Cu in anammox reactors were tracked. • The response of the EPS to Cu(II) was characterized by 3D-EEM spectra. • The mechanism of Cu inhibition on anammox granules was updated. • The feasibilities of two novel remediation strategies were investigated. - Abstract: In this study, the behavior, distribution and form dynamics of overloaded Cu(II) in anaerobic ammonium oxidation (anammox) granular sludge reactors were investigated. The performance and physiological characteristics were tracked by continuous-flow monitoring to evaluate the long-term effects. High Cu loading (0.24 g L{sup −1} d{sup −1}) exceeded sludge bearing capacity, and precipitation dominated the removal pathway. The Cu distribution migrated from the extracellular polymeric substances-bound to the cell-associated Cu and the Cu forms shifted from the weakly bound to strongly bound fractions over time. Pearson correlation and fluorescence spectra analyses showed that the increase in protein concentrations in the EPS was a clear self-defense response to Cu(II) stress. Two remediation strategies, i.e., ethylenediamine tetraacetic acid (EDTA) washing and ultrasound-enhanced EDTA washing, weakened the equilibrium metal partition coefficient from 5.8 to 0.45 and 0.34 L mg{sup −1}SS, respectively, thereby accelerating the external diffusion of the Cu that had accumulated in the anammox granules.

  20. Behavior and fate of copper ions in an anammox granular sludge reactor and strategies for remediation

    International Nuclear Information System (INIS)

    Zhang, Zheng-Zhe; Deng, Rui; Cheng, Ya-Fei; Zhou, Yu-Huang; Buayi, Xiemuguli; Zhang, Xian; Wang, Hui-Zhong; Jin, Ren-Cun

    2015-01-01

    Highlights: • The Cu partition in an anammox UASB reactor was predicted by models. • The distribution and form dynamics of Cu in anammox reactors were tracked. • The response of the EPS to Cu(II) was characterized by 3D-EEM spectra. • The mechanism of Cu inhibition on anammox granules was updated. • The feasibilities of two novel remediation strategies were investigated. - Abstract: In this study, the behavior, distribution and form dynamics of overloaded Cu(II) in anaerobic ammonium oxidation (anammox) granular sludge reactors were investigated. The performance and physiological characteristics were tracked by continuous-flow monitoring to evaluate the long-term effects. High Cu loading (0.24 g L −1 d −1 ) exceeded sludge bearing capacity, and precipitation dominated the removal pathway. The Cu distribution migrated from the extracellular polymeric substances-bound to the cell-associated Cu and the Cu forms shifted from the weakly bound to strongly bound fractions over time. Pearson correlation and fluorescence spectra analyses showed that the increase in protein concentrations in the EPS was a clear self-defense response to Cu(II) stress. Two remediation strategies, i.e., ethylenediamine tetraacetic acid (EDTA) washing and ultrasound-enhanced EDTA washing, weakened the equilibrium metal partition coefficient from 5.8 to 0.45 and 0.34 L mg −1 SS, respectively, thereby accelerating the external diffusion of the Cu that had accumulated in the anammox granules.

  1. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1987-01-01

    This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 10 8 per m 3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

  2. In-reactor corrosion behavior of stainless steel in high temperature sodium

    International Nuclear Information System (INIS)

    Weber, J.W.

    1976-01-01

    Sodium-cladding chemical interaction has been studied on fuel pins clad with 20 percent cold worked type 316 stainless steel and irradiated in the EBR-II at temperatures up to 705 0 C and for exposures to 5300 hours. The measured corrosion rate of the cladding surface immediately above the top of the fuel column was 12.5 μm per year at 690 0 C. The loss of Ni at 700 0 C resulted in the formation of a ferrite layer approximately 5 μm thick. A zone depleted in Ni and Cr extends into the austenite from the ferrite-austenite interface an additional distance of approximately 15 μm. No large changes in volumetric average carbon or nitrogen were observed. The results suggest a slight increase in carbon during the first 2300 hours followed by a slow decrease for longer exposure times. A 25 to 30 percent decrease in bulk nitrogen was observed in cladding with an initial nitrogen content of 225 ppM. Cladding with a lower initial nitrogen content of 63 ppM showed essentially no change. Fuel pin cladding corrosion behavior agrees well with that predicted using out-of-reactor Na loop data, thus providing confidence that out-of-reactor corrosion data can be reliably used in predicting in-reactor fuel cladding performance

  3. The Effects of Internal Components' Disposition on Thermal-Hydraulic Behaviors in Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Han, Ji Woong; Eoh, Jae Hyuk; Kim, Seong O

    2009-01-01

    Decay heat removal is very important in a nuclear power plant. The KALIMER-600, Korea Advanced Liquid MEtal Reactor, employs the PDRC(Passive Decay heat Removal Circuit) to remove the decay heat. However the cooling performance before the activation of DHX greatly depends on the natural circulation flow within the reactor pool. In the previous studies the effect of various design parameters such as coastdown flow, IHX(Intermediate Heat eXchanger) elevation and heat transfer via CCS (Cavity Cooling System) on the initial cooling performance has been analyzed. In the case of IHX elevation analysis the increase of IHX elevation was shown to enhance the initial cooling performance. However, the elevating the IHX is accompanied by the variation of hot or cold pool volume, the previous calculation was resulted from the combination of those effects. In order to analyze those effects qualitatively supplementary calculation conditions were prepared and related analyses have been done in this study. In those analyses the ratio between hot and cold pool volumes has been varied without elevating the IHX by changing the vertical position of separation plate and baffle plate. The COMMIX-1AR/P code is utilized as a tool to investigate overall transient behaviors within a pool. This study is expected to provide the basic information for the decision of internal components' layout in the sodium cooled fast reactor

  4. Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

    International Nuclear Information System (INIS)

    Morimoto, Y.; Madarame, H.; Okamoto, K.

    2001-01-01

    Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

  5. Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.

    1978-06-01

    Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

  6. The behavior of fuel and fission products in manoeuvring the power of the water cooled reactors

    International Nuclear Information System (INIS)

    Luzanova, L.M.; Miglo, V.N.; Slavyagin, P.D.

    1987-01-01

    Problems relating to investigation of dioxide fuel behavior and variation of fission product activity in power manoeuvring in the WWER-type water-cooled reactors are considered. It is pointed out that at loads which do not result in onset of a zone of structural changes the behavior of fuel and that of fission products in the fuel-cladding gap and in the primary coolant under the stationary and transient conditions are determined only by the fuel-cladding interactions. At any power variations the fuel is not subject to structural changes and no excess release of fission products from the fuel occurs. In the presence of developed zones of structural changes in failed fuel elements release of radioactive fission products to the primary coolant is determined only by the thermal regime in the zone of structural changes and is practically independent of the degree of the fuel-cladding interaction. (author). 1 ref., 7 figs

  7. Single Phase Natural Circulation Behaviors of the Integral Type Marine Reactor Simulator under Rolling Motion Condition

    Directory of Open Access Journals (Sweden)

    Hou-jun Gong

    2015-01-01

    Full Text Available During operation in the sea the reactor natural circulation behaviors are affected by ship rolling motion. The development of an analysis code and the natural circulation behaviors of a reactor simulator under rolling motion are described in this paper. In the case of rolling motion, the primary coolant flow rates in the hot legs and heating channels oscillated periodically, and the amplitude of flow rate oscillation was in direct proportion to rolling amplitude, but in inverse proportion to rolling period. The total mass flow rate also oscillated with half the rolling period, and the average total mass flow rate was less than that in steady state. In the natural circulation under a rolling motion, the flow rate oscillations in the hot legs were controlled by the tangential force; however, the mass flow rate oscillations in the total natural circulation and the heating channels were a result of the combined action of the change of inclination angle, flow resistance, and the extra force arising from the rolling motion. The extra tangential force brought about intense flow rate oscillations in the hot legs, which resulted in increasing total flow resistance; however the extra centrifugal force played a role in increasing thermal driving head.

  8. Chaotic behavior in a hydrodynamic model of a fluidized bed reactor

    International Nuclear Information System (INIS)

    Schouten, J.C.; van den Bleek, C.M.

    1991-01-01

    Recent preliminary experimental studies using time-series analysis have demonstrated that the multi-phase flow in fluidized bed reactors can be characterized as chaotic. In the present paper, it is therefore argued that the chaotic time-dependence of fluidization is a characteristic feature which should be included in scaling rules for fluidized bed reactors. For example, the similarity groups applied in dimensionless fluidized bed scaling should be improved by extending them with functions of the relevant numbers from chaos theory, such as the correlation and embedding dimension or the maximum Lyapunov exponent. This requires that the dependence of these numbers on fluidization parameters must be theoretically and experimentally investigated. The concept of chaos in fluidization also requires that the classical, empirically developed, hydrodynamic models that are applied in fluidized bed scaling are amended to include time-dependence, non-linearity as well as a sufficient level of complexity before they can predict any chaotic behavior. An example is given of chaotic behavior generated in the classical counter-current flow model according to Van Deemter by writing the upwards solids velocity as a harmonic oscillating function of time. A low-dimensional strange attractor is found, embedded in two-dimensional phase space, of which the correlation dimension depends on the solids exchange coefficient

  9. Study on membrane reactors for biodiesel production by phase behaviors of canola oil methanolysis in batch reactors.

    Science.gov (United States)

    Cheng, Li-Hua; Yen, Shih-Yang; Su, Li-Sheng; Chen, Junghui

    2010-09-01

    In comparison with the general stirring batch reactor, the membrane reactor has been reported to have higher molar ratios of methanol to oil but ultralow catalyst concentration in the biodiesel production. In this research, the methanolysis of canola oil is conducted in a stirring batch reactor in the presence of NaOH as a catalyst. Based on the investigation of the effects of operating conditions, including methanol to oil molar ration, catalyst concentrations and temperatures, the time course of the reaction path for the reactant composition in the ternary phase diagram of oil-FAME-MeOH offers an effective way to understand the operation of membrane reactors in the biodiesel production. The results show that increasing the residence time of the whole reactant system within the two-phase zone is good for the separation operation through the membranes. (c) 2010 Elsevier Ltd. All rights reserved.

  10. The development of patterns of stable, transient, and school-age onset aggressive behavior in young children.

    Science.gov (United States)

    Kingston, L; Prior, M

    1995-03-01

    To examine the development of patterns of aggressive behavior in children from the age of 2 to 8 years. Children with early histories of aggressive behavior were selected from a community sample of 2,400 infants participating in a longitudinal study. The sample was divided into four groups: children with stable aggressive behavior, those with transient aggression, those with aggression only after age 5 years (late onset), and a comparison group of nonaggressive children. Children with stable aggressive behavior were characterized by a difficult temperament, hostile sibling interactions, maternal perception of the child as difficult, and harsher child-rearing practices. Children whose early aggression decreased over time and those who became aggressive only after entering school could not be reliably classified with the selected family variables. Teacher ratings of temperament factors of task orientation and reactivity and ability ratings correctly classified 74% of children whose aggression began at school-age. Children with persistent aggressive behavior differed from those who improved, predominantly in terms of symptom severity. Problems with aggression can be identified early in development, and a significant proportion of aggressive children are at risk for continuing social and scholastic difficulties. Knowledge of associated factors may play an important role in prevention.

  11. Numerical investigation on stress corrosion cracking behavior of dissimilar weld joints in pressurized water reactor plants

    Directory of Open Access Journals (Sweden)

    Lingyan Zhao

    2014-07-01

    Full Text Available There have been incidents recently where stress corrosion cracking (SCC observed in the dissimilar metal weld (DMW joints connecting the reactor pressure vessel (RPV nozzle with the hot leg pipe. Due to the complex microstructure and mechanical heterogeneity in the weld region, dissimilar metal weld joints are more susceptible to SCC than the bulk steels in the simulated high temperature water environment of pressurized water reactor (PWR. Tensile residual stress (RS, in addition to operating loads, has a great contribution to SCC crack growth. Limited experimental conditions, varied influence factors and diverging experimental data make it difficult to accurately predict the SCC behavior of DMW joints with complex geometry, material configuration, operating loads and crack shape. Based on the film slip/dissolution oxidation model and elastic-plastic finite element method (EPFEM, an approach is developed to quantitatively predict the SCC growth rate of a RPV outlet nozzle DMW joint. Moreover, this approach is expected to be a pre-analytical tool for SCC experiment of DMW joints in PWR primary water environment.

  12. Effect of influent COD/SO4(2-) ratios on biodegradation behaviors of starch wastewater in an upflow anaerobic sludge blanket (UASB) reactor.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Ni, Jialing; Hojo, Toshimasa; Kubota, Kengo; Li, Yu-You

    2016-08-01

    A lab-scale upflow anaerobic sludge blanket (UASB) has been run for 250days to investigate the influence of influent COD/SO4(2-) ratios on the biodegradation behavior of starch wastewater and process performance. Stepwise decreasing COD/SO4(2-) ratio enhanced sulfidogenesis, complicating starch degradation routes and improving process stability. The reactor exhibited satisfactory performance at a wide COD/SO4(2-) range ⩾2, attaining stable biogas production of 1.15-1.17LL(-1)d(-1) with efficient simultaneous removal of total COD (73.5-80.3%) and sulfate (82.6±6.4%). Adding sulfate favored sulfidogenesis process and diversified microbial community, invoking hydrolysis-acidification of starch and propionate degradation and subsequent acetoclastic methanogenesis; whereas excessively enhanced sulfidogenesis (COD/SO4(2-) ratios UASB technology in water industry from basic science. Copyright © 2016 Elsevier Ltd. All rights reserved.

  13. Procedures for the determination of stable elements in construction materials from the nuclear reactors at Risoe National Laboratory

    International Nuclear Information System (INIS)

    Oestergaard, L.F.

    2006-03-01

    Methods for the accurate determination of stable isotopes of elements in construction materials with relevance to the work of the Danish Decommissioning have been developed. Prior to the analysis the elements of interest must be released from the construction materials and this is done with several different digestion methods. For the analysis of aluminium, lead, graphite and steels the samples are digested with mineral acids and microwave heating at increased pressures in a sealed teflon vessel. The aluminium, lead and steel are fully dissolved after the digestion procedure whereas graphite is chemically inert to the acid treatment used, but the elements of interest are extracted from the graphite quite efficiently. Concrete is digested with open-vessel heating in a Modblock TM digesting unit in a two step procedure involving 40% HF followed by 32% HNO 3 . The heavy barite concrete is first treated as the concrete samples but a large residue of poorly soluble sulphates (mainly BaSO 4 ) is left. The residue is fused with NaOH/Na 2 CO 3 at 575 deg. C and after some work up the product from the fusion is dissolved in dilute HNO 3 . After the release of the elements from the materials, the samples are analysed by ICP-OES and ICP-MS multi-element analysis. In general the following elements are of interest to DD; Ag, Ba, Ca, Co, Eu, Fe, Li, Mo, Nb, Ni, Sm, Th and U. For graphite, steel, concrete and heavy concrete, analytical methods for the determination of all 13 elements have been developed (except Ca in steel). For aluminium and lead methods for the determination of Ag, Co, Li, Nb, Ni and U, as well as Ba in the lead have been developed. When possible the methods have been verified against certified reference materials and calibration with standards additions and internal standard corrections have been used to correct for matrix effects most efficiently. The accuracy has also been checked with spikes when reference materials are not available. For the aluminium

  14. Procedures for the determination of stable elements in construction materials from the nuclear reactors at Risoe National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Oestergaard, L.F.

    2006-03-15

    Methods for the accurate determination of stable isotopes of elements in construction materials with relevance to the work of the Danish Decommissioning have been developed. Prior to the analysis the elements of interest must be released from the construction materials and this is done with several different digestion methods. For the analysis of aluminium, lead, graphite and steels the samples are digested with mineral acids and microwave heating at increased pressures in a sealed teflon vessel. The aluminium, lead and steel are fully dissolved after the digestion procedure whereas graphite is chemically inert to the acid treatment used, but the elements of interest are extracted from the graphite quite efficiently. Concrete is digested with open-vessel heating in a Modblock{sup TM} digesting unit in a two step procedure involving 40% HF followed by 32% HNO{sub 3}. The heavy barite concrete is first treated as the concrete samples but a large residue of poorly soluble sulphates (mainly BaSO{sub 4}) is left. The residue is fused with NaOH/Na{sub 2}CO{sub 3} at 575 deg. C and after some work up the product from the fusion is dissolved in dilute HNO{sub 3}. After the release of the elements from the materials, the samples are analysed by ICP-OES and ICP-MS multi-element analysis. In general the following elements are of interest to DD; Ag, Ba, Ca, Co, Eu, Fe, Li, Mo, Nb, Ni, Sm, Th and U. For graphite, steel, concrete and heavy concrete, analytical methods for the determination of all 13 elements have been developed (except Ca in steel). For aluminium and lead methods for the determination of Ag, Co, Li, Nb, Ni and U, as well as Ba in the lead have been developed. When possible the methods have been verified against certified reference materials and calibration with standards additions and internal standard corrections have been used to correct for matrix effects most efficiently. The accuracy has also been checked with spikes when reference materials are not

  15. Stable isotopes reveal rail-associated behavior in a threatened carnivore.

    Science.gov (United States)

    Hopkins, John B; Whittington, Jesse; Clevenger, Anthony P; Sawaya, Michael A; St Clair, Colleen Cassady

    2014-01-01

    Human-wildlife conflict is a leading cause of adult mortality for large carnivores worldwide. Train collision is the primary cause of mortality for threatened grizzly bears (Ursus arctos) in Banff National Park. We investigated the use of stable isotope analysis as a tool for identifying bears that use the railway in Banff. Rail-associated bears had higher δ(15)N and δ(34)S values than bears sampled away from the rail, but similar δ(13)C values. Because elevated δ(15)N values are indicative of higher animal protein consumption, rail-associated bears likely preyed on ungulates that foraged along the rail or scavenged on train-killed animals. The higher δ(34)S values in bear hair could have resulted from bears consuming sulfur pellets spilled on the rail or through the uptake of sulfur in the plants bears or animals consumed. Similar δ(13)C values suggest that the two types of bears had generally similar plant-based diets. Results from this study suggest that stable isotopes analysis could be used as a non-invasive, affordable, and efficient technique to identify and monitor bears that forage on the railway in Banff and potentially other transportation corridors worldwide.

  16. Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor

    International Nuclear Information System (INIS)

    Yang Zhimin

    2000-01-01

    The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)

  17. Theoretical and experimental study on the magnetomechanical behavior of superconducting helical coils for a fusion reactor

    International Nuclear Information System (INIS)

    Takaghi, T.; Miya, K.; Yamada, H.; Takagi, T.

    1984-01-01

    The magnetomechanical behavior of superconducting helical coils for a magnetic fusion reactor was investigated experimentally and theoretically. Deformations of straight and torus type helical coils were caused due to static electromagnetic forces in the liquid helium cryostat and were analysed with the finite element computer code made here. Despite of a large scatter of experimental data due to a non-uniform friction force between the helical coil and the torus of stainless steel, the numerical results are very close to the mean value of the data. Numerical analysis of the force distribution acting on the helical coils was also performed for a Heliotron's coil system to characterize its nature. The force could be categorized conveniently as an extensional force, a tangential force and a toroidal force which correspond respectively to the kind of forces acting on toroidal field coils. Additionally, the effect of mechanical constraint on the magnetomechanical behavior is discussed and shows that the location of the constraint significantly affects the stress distributions in the coils. (orig.)

  18. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland

    2002-09-30

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  19. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland, Ph.D.

    2000-01-01

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  20. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland, Ph.D.

    2002-01-01

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors.

  1. Dynamics and control of molten-salt breeder reactor

    Directory of Open Access Journals (Sweden)

    Vikram Singh

    2017-08-01

    Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  2. Comparison of the transient behavior of lead-based advanced critical and sub-critical reactors

    International Nuclear Information System (INIS)

    Wang Gang; Gu Zhixing; Wang Zhen; Jin Ming; Bai Yunqing

    2014-01-01

    A lead-based reactor developed by FDS Team is proposed in 2011 and designed to be 10 MW. It is a pool type reactor and the primary coolant is driven by natural circulation. The reactor has two operation modes, which are a lead-based critical fast reactor mode and a lead-based sub-critical reactor mode. The conceptual designs of the two modes are both completed by 2013. In this paper, four transient accidents were simulated for both the critical and sub-critical reactors above by NTC-2D code, which is developed by FDS Team for advanced reactor safety analysis. The four accidents were protected and unprotected loss of heat sink accidents (PLOHS and ULOHS), protected and unprotected transient overpower accidents (PTOP and UTOP). The simulation results of the two reactors were compared and analyzed. The results showed that during PLOHS and PTOP accidents for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the sub-critical reactor increased bigger than those of the critical one. For UTOP, the parameters above of the critical fast reactor were much bigger than those of the sub-critical one. The analysis results showed different safety advantages of the lead-based critical fast and sub-critical reactors during different transient accidents. (author)

  3. Effect of hydraulic retention time on hydrodynamic behavior of anaerobic-aerobic fixed bed reactor treating cattle slaughterhouse effluent

    Directory of Open Access Journals (Sweden)

    Daiane Cristina de Freitas

    2017-09-01

    Full Text Available The study of the hydrodynamic behavior in reactors provides characteristics of the flow regime and its anomalies that can reduce biological processes efficiency due to the decrease of the useful volume and the hydraulic retention time required for the performance of microbial activity. In this study, the hydrodynamic behavior of an anaerobic-aerobic fixed bed reactor, operated with HRT (hydraulic retention time of 24, 18 and 12 hours, was evaluated in the treatment of raw cattle slaughterhouse wastewater. Polyurethane foam and expanded clay were used as support media for biomass immobilization. Experimental data of pulse type stimulus-response assays were performed with eosin Y and bromophenol blue, and adjusted to the single-parameter theoretical models of dispersion and N-continuous stirred tank reactors in series (N-CSTR. N-CSTR model presented the best adjustment for the HRT and tracers evaluated. RDT (residence time distribution curves obtained with N-CSTR model in the assays with bromophenol blue resulted in better adjustment compared to the eosin Y. The predominant flow regime in AAFBR (anaerobic aerobic fixed bed reactor is the N-CSTR in series, as well as the existence of preferential paths and hydraulic short-circuiting.

  4. Synthesis of preliminary studies on the mechanical behavior of RJH reactor fuel

    International Nuclear Information System (INIS)

    Witters, F.; Mori, V.; Guigon, B.

    1999-01-01

    The experimental Reactor Jules Horowitz (RJH) will succeed to OSIRIS. It will have to satisfy the needs for technological irradiation necessary for further developments in Pressurized Water Reactors and those associated with research on new reactor systems. These objectives require a broad range of operating conditions, which the evolutionary nature of the core will make possible. The maximum fast flux in the core will be about 5 10 14 n cm 2 s -1 . As this level was never reached with silicide fuel plates, an investigation of their behaviour was conducted and is presented here. Beforehand, the material testing reactor fuels are briefly described. (authors)

  5. Behavior of radon, chemical compounds and stable elements in underground water

    International Nuclear Information System (INIS)

    Lopez R, N.; Segovia, N.; Lopez, M.B.E.; Pena, P.; Armienta, M.A.; Godinez, L.; Seidel, J.L.

    2001-01-01

    The radon behavior, chemical compounds, major and trace elements in water samples of four springs and three wells of urban and agricultural zones around the Jocotitlan volcano and El Oro region was determined, both of them located in the medium part of the Mexican neo-volcanic axis. The 222 Rn was measured by the liquid scintillation method, the analysis of major components was realized with conventional chemical techniques, while the trace elements were quantified using an Icp-Ms. The average values of the radon concentrations obtained during one year were constant relatively, in an interval from 0.97 to 4.99 Bq/lt indicating a fast transport from the reload area toward the sampling points. the compounds, major and trace elements showed differences which indicate distinct origins of water from the site studies. (Author)

  6. Cluster analysis to evaluate stable chemical elements and physical-chemical parameters behavior on uranium mining waste

    International Nuclear Information System (INIS)

    Pereira, Wagner de Souza; Py Junior, Delcy de Azevedo; Goncalves, Simone; Kelecom, Alphonse; Morais, Gustavo Ferrari de; Campelo, Emanuele Lazzaretti Cordova; Dores, Luis Augusto de Carvalho Bresser

    2011-01-01

    The Ore Treating Unit (UTM, in portuguese) is a deactivated uranium mine. A cluster analysis was used to evaluate the behavior of stable chemical elements and physical-chemical parameters in their effluents. The utilization of the cluster analysis proved itself effective in the assessment, allowing the identification of groups of chemical elements, physical-chemical parameters and their joint analysis (elements and parameters). As a result we may assert, based on data analysis, that there is a strong link between calcium and magnesium and between aluminum and rare-earth oxides on UTM's effluents. Sulphate was also identified as strongly linked to total and dissolved solids, and those to electrical conductivity. There were other associations, but not so strongly linked. Further gathering, to seasonal evaluation, are required in order to confirm those analysis. Additional statistical analysis (factor analysis) must be used to try to identify the origin of the identified groups on this analysis. (author)

  7. Dynamic behavior of large oxide-fueled fast reactors during over-power transients due to boiling of sodium

    International Nuclear Information System (INIS)

    Pires, L.F.G.

    1983-01-01

    The dynamic behavior of large oxide-fueled fast reactors during over-power transients or under-flow situations which result in boiling of sodium used as coolant. The fuel heat transfer was analysed to determine the fuel temperature profile and the Doppler feedback reactivity. The sodium pressure, temperature, mass flow rate and sodium voiding reactivity were obtained by solving the basic coolant hydrodynamics equations. (Author) [pt

  8. Load bearing capacities and elastic-plastic behavior of reactor vessel internals

    International Nuclear Information System (INIS)

    Watanabe, Keita; Nagase, Ryuichi

    2017-01-01

    Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

  9. Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

    1978-05-01

    Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: /sup 58/Ni + n ..-->.. /sup 59/Ni + ..gamma..; /sup 59/Ni + n ..-->.. /sup 56/Fe + ..cap alpha... Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case.

  10. Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

    International Nuclear Information System (INIS)

    Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

    2015-01-01

    Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

  11. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1993-01-01

    Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

  12. Experimental simulation study on hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Zhang Youjie; Jia Haijun; Bo Jinhai; Hong Liuming; Bo Hanliang; Liu Zhiyong

    1997-07-01

    The hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor is studied through a 1:2.33 test model. The design and other feature of the test model is described. The experimental results show that the flow resistance coefficient of the heat exchanger becomes self-simulation when Reynolds number is greater than 5000. The value of flow resistance coefficient at self-simulation condition and the distribution of pressure drop in the heat exchanger are given through experiment. The option design to reduce flow resistance is proposed. The designed and experimental value for the flow resistance coefficient are in good agreement. The variation of system parameters during flow excursion was described. The experimental results are of great significant for the final design of the main heat exchanger of Daqing 200 MW nuclear heating reactor. (2 refs., 5 figs., 1 tab.)

  13. Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Alhabet, F.

    2004-02-01

    Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed

  14. Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

    International Nuclear Information System (INIS)

    Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

    1999-08-01

    The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

  15. Seismic response analysis of nuclear reactor buildings under consideration of soil-structure interaction with torsional behavior

    International Nuclear Information System (INIS)

    Mizuno, N.; Iida, T.; Tsushima, Y.; Araki, T.; Nojima, O.

    1977-01-01

    In this paper, the seismic response analysis is described in detail for estimating the soil-structure interaction effects with the torsional behavior. The analytical method is firstly shown for estimating the stiffness of reactor building by the bending-shear and torsion theory of the thin wall sections in regard to the behavior of structure. The three-dimensional behavior of structure can be obtained more briefly and simply by the proposed method. Secondly, the dynamical soil-foundation coefficient for estimating the dissipation of vibrational energy on the ground is derived by H. Tajimi's theory which is based on a solution of the propagation of seismic waves caused by point excitation on the surface of the elastic half-space medium. The above results give the vibrational impedances of the soil-foundation corresponding to the static soil coefficient, which is defined to the excitation force in the frequency domain. In order to analyze to the equivalues of reactor building, the authors thirdly attempt to approximate the dynamic soil-foundation coefficient as the frequency transfer function of displacement. The complex damping is used for more suitably estimating the elastic structural damping effects of structure. The regression analysis of many degrees of freedom is fourthly attempted for estimating the natural periods annd equivalent viscous damping ratios directly from the experimental results by the forced vibrational test performed in 1974. The analytical results are finally shown for simulating and comparing with the above-mentioned experimental results

  16. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1992-01-01

    Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements

  17. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

  18. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

  19. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.

    1987-01-01

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author) [pt

  20. Comparison of oxide- and metal-core behavior during CRBRP [Clinch River Breeder Reactor Plant] station blackout

    International Nuclear Information System (INIS)

    Polkinghorne, S.T.; Atkinson, S.A.

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs

  1. Thermohydraulic behavior in a primary cooling system during a loss-of-coolant accident of a light-water reactor

    International Nuclear Information System (INIS)

    Shimamune, Hiroji; Shiba, Masayoshi; Adachi, Hiromichi; Suzuki, Norio; Okubo, Kaoru

    1975-12-01

    With ROSA-I (Rig of Safety Assessment - I), 61 runs of the LWR blowdown experiment have been carried out under the conditions: model reactor type, BWR and PWR; reactor core, none, no-heating and heating; rupture position, upper and lower pressure vessel nozzle; initial discharge pressure, 40, 70 and 100 kg/cm 2 G; and rupture diameter, 25, 50, 70, 100 and 125 mm. The purpose was to obtain the data of thermal and hydrodynamic behavior in the reactor pressure vessel during a blowdown, including in-vessel pressure, coolant temperature, discharge flow rate, model fuel rod surface temperature and shock wave. Analysis was also made with the codes RELAP-2 and -3 developed by NRTS of the United States, to verify the calculation model used. In addition, the results of calculation with the shockwave analysis code DEPCO developed in JAERI were compared with those by experiment. The experimental facility ROSA-I and the results obtained with it and also the analyses made in this connection, are described in detail. (auth.)

  2. Effects of Cognitive Behavioral Therapy for Insomnia on Sleep-Related Cognitions Among Patients With Stable Heart Failure.

    Science.gov (United States)

    Redeker, Nancy S; Jeon, Sangchoon; Andrews, Laura; Cline, John; Mohsenin, Vahid; Jacoby, Daniel

    2017-07-26

    Cognitive behavioral therapy for insomnia (CBT-I) improves insomnia and fatigue among chronic heart failure (HF) patients, but the extent to which sleep-related cognitions explain CBT-I outcomes in these patients is unknown. We examined the effects of CBT-I on sleep-related cognitions, associations between changes in sleep-related cognitions and changes in sleep and symptoms after CBT-I, and the extent to which cognitions mediated the effects of CBT-I. Stable New York Heart Association Class II-III HF patients (total n = 51; n = 26 or 51.0% women; M age = 59.1 ± 15.1 years). HF patients were randomized in groups to group CBT-I (n = 30) or attention control (HF self-management education, n = 21) and completed actigraphy, the Insomnia Severity Index, Pittsburgh Sleep Quality Index, Dysfunctional Beliefs and Attitudes about Sleep (DBAS) and Sleep Disturbance Questionnaires (SDQ), and self-reported fatigue, depression, anxiety, and sleepiness (baseline, immediately after treatment, six months). We used mixed-effects modeling, mediation analysis with a bootstrapping approach, and Pearson correlations. There was a statistically significant group × mult time effect on DBAS. DBAS mediated the effects of CBT-I on insomnia severity and partially mediated CBT-I effects on fatigue. Improvements in dysfunctional cognitions were associated with improved sleep quality, insomnia severity, sleep latency and decreased fatigue, depression, and anxiety, with sustained effects at six months. Improvement in dysfunctional sleep-related cognitions is an important mechanism for CBT-I effects among HF patients who are especially vulnerable to poor sleep and high symptom burden.

  3. Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    2001-01-01

    The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff ∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early

  4. Co-pyrolysis behaviors of saw dust and Shenfu coal in drop tube furnace and fixed bed reactor.

    Science.gov (United States)

    Li, Shuaidan; Chen, Xueli; Wang, Li; Liu, Aibin; Yu, Guangsuo

    2013-11-01

    Co-pyrolysis behaviors of saw dust (SD) and Shenfu bituminous coal (SF) were studied in a drop tube furnace and a fixed bed reactor at different temperatures respectively. Six different biomass/coal ratios (B:C) were used. Compared the results with the calculated value obtained by the additional behavior, CO volume yields were lower while H2, CH4, CO2, volume yields were higher. Blend char yields had a good agreement with the calculated values, and their structures remained similar with SD and SF char's. Synergy effect occurred in gaseous phase, which was mainly caused by the secondary reactions. Compared the blend char yields in the drop tube furnace with those in the fixed bed reactor, the results showed the contacting way of biomass and coal particles had little influence on char yield in co-pyrolysis process. The reactivity index of blend char achieved the minimum at B:C=40:60 and the maximum at B:C=80:20. Copyright © 2013 Elsevier Ltd. All rights reserved.

  5. Biomass Gasification Behavior in an Entrained Flow Reactor: Gas Product Distribution and Soot Formation

    DEFF Research Database (Denmark)

    Qin, Ke; Jensen, Peter Arendt; Lin, Weigang

    2012-01-01

    Biomass gasification and pyrolysis were studied in a laboratory-scale atmospheric pressure entrained flow reactor. Effects of operating parameters and biomass types on the syngas composition were investigated. In general, the carbon conversion during biomass gasification was higher than 90......% at the optimal conditions of 1400 °C with steam addition. The biomass carbon that was not converted to gas in the gasification process only appeared as soot particles in the syngas in all of the experiments, except for the two experiments performed at 1000 °C, where a very small amount of char was also left....... In comparison to pyrolysis, lower yields of soot, H2, and CO were produced during gasification. The yield of soot could be reduced by a longer residence time, larger feeder air flow, lower oxygen concentration, higher excess air ratio, higher steam/carbon ratio, and higher reactor temperature. Changes...

  6. Actinide behavior in the integral fast reactor. Progress report, May 1, 1992--April 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1993-05-01

    Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory`s site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core.

  7. Mass transfer behavior of rotating square cylinder electrochemical reactor in relation to wastewater treatment

    International Nuclear Information System (INIS)

    Abdel-Aziz, M.S.M.; El-Shazly, A.H.; Farag, H.A.; Sedahmed, G.H.

    2011-01-01

    Highlights: → The work explores a new electrochemical reactor by using square rotating cylinders. → The results show that it is superior to the traditional circular rotating cylinder. → A dimensionless design equation for the new reactor was correlated. → The oxalic acid removal by the new reactor was succeeded and found promising. → The energy consumption per kg oxalic acid removed by the unit was calculated. - Abstract: Rates of mass transfer at a rotating square cylinder were measured by an electrochemical technique which involved measuring the limiting current of the cathodic reduction of K 3 Fe(CN) 6 in a large excess of NaOH solution. Variables studied were: cylinder rotation speed, physical properties of the solution and cylinder equivalent diameter. The data for the condition 1577 0.33 Re 0.45 For a given set of conditions the rate of mass transfer at the square rotating cylinder was found to be higher than that at the traditional circular rotating cylinder by an amount ranging from 47% to 200% depending on Re. The use of the square rotating cylinder electrode in removing oxalic acid from wastewater by anodic oxidation on Pb/PbO anode was examined and found to be promising.

  8. The behavior of 131I in polymetatelluric acid irradiated in the nuclear reactor

    International Nuclear Information System (INIS)

    Teofilovski, C.

    1966-01-01

    Polymetarelluric acid, whose composition is (H 2 TeO 4 ) n , is successfully used at he Institute as a target for obtaining 131 I in the reactor. It is prepared by hearing orthotelluric acid in air at 160 deg C of in a steam of water vapor at 208 deg C. Analysis of the valency states of 131 I in irradiated (H 2 TeO 4 ) n prepared in either of the above ways shows a variable ratio of reduced and oxidized forms. A considerable increase of the reduced forms with increasing integral thermal neutron flux during irradiation in the reactor in the given interval has also been observed. In order to explain the above phenomenon (H 2 TeO 4 ) n was irradiated in the reactor under different conditions, with measurement of the wall temperature of the quartz ampoules containing the target material. Yields of reduced and oxidized form of 131 I were determined immediately after irradiation and after annealing of the target at temperatures from 60 deg C to 150 deg C. A considerable decrease in the yield of the reduced forms of 131 I on target annealing above 100 deg C was observed (author)

  9. Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

    Directory of Open Access Journals (Sweden)

    Ordóñez Ródenas José

    2016-01-01

    Full Text Available One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included.

  10. Modeling transient thermal hydraulic behavior of a thermionic fuel element for nuclear space reactors

    International Nuclear Information System (INIS)

    Al-Kheliewi, A.S.; Klein, A.C.

    1994-01-01

    A transient code (TFETC) for determining the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) during changes in operating conditions has been successfully developed and tested. A fully implicit method is used to solve the system of equations for temperatures at each time step. Presently, TFETC has the ability to handle the following transients: startup, loss of flow accidents, and shutdown. The code has been applied to the startup of the ATI single cell configuration which appears to start up and shut down in an orderly and reasonable fashion. No unexpected transient features were observed. The TFE also appears to function robustly under loss of flow accident conditions. It appears hat sufficient time is available to shut the reactor down safely without melting point the fuel. The model shows that during a complete loss of flow accident (without shutdown) the coolant reaches its boiling point in approximately 35 seconds. The fuel may exceed its melting point after this time as the NaK coolant will boil if the reactor is not shut down. For 1/2, 1/3, and 1/4 pump failures, the fuel temperatures never exceed the fuel melting point even if the reactor is not shut down

  11. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  12. The Behavior of Pilot Trickle-Bed Reactor under Periodic Operation

    Czech Academy of Sciences Publication Activity Database

    Tukač, V.; Šimíčková, M.; Chyba, V.; Lederer, J.; Kolena, J.; Hanika, Jiří; Jiřičný, Vladimír; Staněk, Vladimír; Stavárek, Petr

    2007-01-01

    Roč. 62, 18-20 (2007), s. 4891-4895 ISSN 0009-2509. [International Symposium on Chemical Reaction Engineering - From Science to Innovative Engineering /19./. Potsdam/Berlin, 03.09.2006-06.09.2006] R&D Projects: GA MPO(CZ) FT-TA/039 Grant - others:CYCLOP(XE) G1RD/CT2000/00225 Institutional research plan: CEZ:AV0Z40720504 Source of funding: R - rámcový projekt EK Keywords : olefine hydrogenation * pilot-scale * trickle-bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 1.775, year: 2007

  13. Mechanical behavior of a fast reactor core: Application of the 3D codes to SUPER PHENIX 1

    International Nuclear Information System (INIS)

    Bernard, A.; Masoni, P.; Dorsselaere, J.P. van

    1983-01-01

    The series of the 3-dimensional mechanical codes of a fast reactor core was used for the first time within the framework of a design study of an industrial reactor: SUPER-PHENIX 1. These codes are the following ones: - ARGOH which calculates the behavior of an isolated subassembly. - HARMONIE which calculates the core mechanical equilibrium - TRACAR which yields a graphic visualization of HARMONIE results, and calculates the handling forces and support reactions - HARMOREA which calculates the reactivity variations between given equilibrium states (for instance: pads effect and diagrid effect); now at the end of its development. The calculations were performed on 1/3 of the SPX1 core. Their purpose is double: - on the one hand, to check design criteria, and provide the loadings for the subassembly mechanical design studies; on the other hand, to evaluate the reactivity effects, related to the horizontal core deformations, and useful for operation and safety studies. The results of these calculations showed that the design criteria were verified for the contractual lifetime of the subassemblies. (orig.)

  14. Iodine stress corrosion cracking of Zircaloy: Laboratory data, a phenomenological model, and predictions of in-reactor behavior

    International Nuclear Information System (INIS)

    Miller, A.K.; Tasooji, A.

    1981-01-01

    Data from laboratory tests performed on unirradiated and irradiated Zircaloy have been used as the basis for developing a phenomenological model of iodine-induced stress corrosion crack initiation and growth. The model is capable of predicting the response of cladding subjected to complex loading conditions. Major features of the data incorporated into the model include the existence of a threshold stress, the effect of iodine concentration, temperature effects, the role of chemical inhomogeneities and mechanical flaws, crack initiation in smooth specimens, crack propagation rates as a function of stress intensity in flawed specimens, and the detrimental effect of irradiation. The major physical processes addressed by the model include intergranular stress corrosion cracking (SCC), transgranular SCC, ductile rupture, iodine penetration by surface diffusion along existing on incipient cracks, and stress and strain intensification and triaxiality caused by cracks or flaws. A probabilistic description of the size distribution of the flaws found in as-fabricated cladding is used at the basis for quantitatively extrapolating the laboratory test results to predict in-reactor cladding behavior. The in-reactor SCC resistance of a large fuel assembly is predicted to be substantially lower and more variable than that of small laboratory specimens. (orig.)

  15. Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

    International Nuclear Information System (INIS)

    Spencer, B.W.; Marchaterre, J.F.

    1985-01-01

    Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core (although reactor-material experiments are needed to confirm these high condensation rates). If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the implication is that the ability of vapor expansion to perform appreciable work on the system is largely eliminated. Furthermore, the ability of an expanding vapor bubble to transport fuel and fission product species to the cover gas region where they may be released to the containment is also largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-15

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

  17. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  18. Exploring the negative temperature coefficient behavior of acetaldehyde based on detailed intermediate measurements in a jet-stirred reactor

    KAUST Repository

    Tao, Tao

    2018-03-20

    Acetaldehyde is an observed emission species and a key intermediate produced during the combustion and low-temperature oxidation of fossil and bio-derived fuels. Investigations into the low-temperature oxidation chemistry of acetaldehyde are essential to develop a better core mechanism and to better understand auto-ignition and cool flame phenomena. Here, the oxidation of acetaldehyde was studied at low-temperatures (528–946 K) in a jet-stirred reactor (JSR) with the corrected residence time of 2.7 s at 700 Torr. This work describes a detailed set of experimental results that capture the negative temperature coefficient (NTC) behavior in the low-temperature oxidation of acetaldehyde. The mole fractions of 28 species were measured as functions of the temperature by employing a vacuum ultra-violet photoionization molecular-beam mass spectrometer. To explain the observed NTC behavior, an updated mechanism was proposed, which well reproduces the concentration profiles of many observed peroxide intermediates. The kinetic analysis based on the updated mechanism reveals that the NTC behavior of acetaldehyde oxidation is caused by the competition between the O-addition to and the decomposition of the CHCO radical.

  19. The Behavior of Palladium as a Getter for Lanthanide Fission Products in U-Mo-Ti-Zr Fast Reactor Fuels

    Science.gov (United States)

    Howard, Cameron Tyler

    One of the hurdles to extending the life of metallic fast reactor fuel alloys is Fuel-Clad Chemical Interaction (FCCI), a phenomenon which occurs between fuel and cladding resulting in thinning of the cladding. The cause of FCCI is the reaction between cladding constituents (e.g. iron and nickel) and lanthanide fission products generated in the fuel (e.g. lanthanum and cerium). This interaction can produce localized melting of the cladding, reducing its thickness over the life of the fuel element. It has been suggested that FCCI can be hindered by doping the fuel with palladium, a candidate getter for lanthanide fission products. There is therefore interest in demonstrating the efficacy of this particular lanthanide getter for realistic fast reactor fuel analogues. Work is presented based on the U-M (M = 50Mo-43Ti-7Zr, wt. pct.) alloy system both with, and without, palladium additions. The research was conducted using depleted uranium alloys developed as metallurgical surrogates for real spent fuels. Burnup was simulated using cerium as a mock lanthanide fission product to assess the behavior of palladium with respect to fuel and cladding constituents. The behavior of palladium in terms of microstructural evolution was studied from both as-cast and annealed surrogate fuel specimens as well as diffusion couples between surrogate fuel alloys and type HT-9 stainless steel cladding. Results derived from characterization of these metallurgical surrogate experiments are presented and it is shown that palladium is a promising getter for lanthanide fission products in the given alloy system.

  20. Study on the Modeling Method for Computational Fluid Dynamics to Predict Coolant Flow Behavior in a Tank-in-pool Type Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Chae, Hee Taek; Park, Cheol; Kim, Heon Il; Lee, Doo Jeong; Seo, Kyoung Woo; Oh, Jae Min; Yoon, Ju Hyeon; Ha, Jae Joo

    2009-08-15

    This is a study on the modeling method for computational fluid dynamics (CFD) to analyze and to predict primary coolant flow behavior in a tank-in-pool type research reactor at the conceptual design stage. It is very hard to conduct a CFD analysis on an enormous complex equipment integrated system like a nuclear reactor with no simplification and alternative model treatment, so it is required a new methodological approach to be adopted to such an enormous CFD model. In this study it is proposed a new modeling method to solve a nuclear reactor employing plate type nuclear fuel assemblies. The idea attributes fuel assembly modeling to simplification, but the effect and the influence to a flow by fuel assembly can be reproduced as close as possible to physical phenomena. The idea is applied to CFD analysis model for a tank-in-pool type research reactor. A careful evaluation of the CFD analysis result is carried out whether the CFD model can provide a reasonable prediction or not. According to the evaluation it can be said that the proposed modeling method is so useful idea for the CFD analysis model to solve the whole nuclear reactor. The CFD result provides valuable predictions to the flow phenomena in the tank-in-pool type research reactor, which contributes to the more understanding of flow characteristics of primary coolant flow in the reactor. It is shown by the CFD analysis that the reactor core would have even flow distribution within {+-}3% deviation to the design flow rate, and that the primary coolant of PCHT(Primary Coolant Hold-up Tank) would flow out to reactor pool via the NC flow gates. The outstanding grasping of the flow in the PCHT leads to a remedy to reduce the flowing out from the NC flow gates.

  1. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core

    International Nuclear Information System (INIS)

    Belo, Allan Cavalcante

    2016-01-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4 th generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range and

  2. The Phase Behavior Effect on the Reaction Engineering of Transesterification Reactions and Reactor Design for Continuous Biodiesel Production

    Science.gov (United States)

    Csernica, Stephen N.

    transitions from two phases to a single phase, or pseudo-single phase. The transition to a single phase or pseudo-single phase is a function of the methanol content. Regardless, the maximum observed reaction rate occurs at the point of the phase transition, when the concentration of triglycerides in the methanol phase is largest. The phase transition occurs due to the accumulation of the primary product, biodiesel methyl esters. Through various experiments, it was determined that the rate of the triglyceride mass transfer into the methanol phase, as well as the solubility of triglycerides in methanol, increases with increasing methyl ester concentration. Thus, there exists some critical methyl ester concentration which favors the formation of a single or pseudo-single phase system. The effect of the by-product glycerol on the reaction kinetics was also investigated. It was determined that at low methanol to triglyceride molar ratios, glycerol acts to inhibit the reaction rate and limit the overall triglyceride conversion. This occurs because glycerol accumulates in the methanol phase, i.e. the primary reaction volume. When glycerol is at relatively high concentrations within the methanol phase, triglycerides become excluded from the reaction volume. This greatly reduces the reaction rate and limits the overall conversion. As the concentration of methanol is increased, glycerol becomes diluted and the inhibitory effects become dampened. Assuming pseudo-homogeneous phase behavior, a simple kinetic model incorporating the inhibitory effects of glycerol was proposed based on batch reactor data. The kinetic model was primarily used to theoretically compare the performance of different types of continuous flow reactors for continuous biodiesel production. It was determined that the inhibitory effects of glycerol result in the requirement of very large reactor volumes when using continuous stirred tank reactors (CSTR). The reactor volume can be greatly reduced using tubular style

  3. 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

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

    International Nuclear Information System (INIS)

    Merk, B.; Weiß, F.P.

    2011-01-01

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

  5. Thermomechanical behavior of fuel particles in a matrix during reactor power excursions

    International Nuclear Information System (INIS)

    Brittan, R.O.; Smith, R.S.

    1977-01-01

    This work determines the largest particle size that can be used in fabricating fuel material without exceeding temperature or stress criteria during transient operation. To do this temperature distribution histories must be determined for various particle sizes and volume fractions using typical power densities histories of transient reactor operation. From these, the critical stresses are calculated. The model chosen to accomplish this is a spherical fuel particle in a spherical matrix shell. Heat flow and temperature continuity conditions are imposed at the interface, and a zero temperature gradient is specified at the outer radius of the matrix shell. The particle power density is assumed to be uniform radially. Provisions are made for uniform power density in the matrix to model gamma heating and power density in interface layers to allow for radiant and fission fragment heating. A computer code was prepared to solve the model performance, yielding the temperature and stress distribution histories. Material property variation with temperature is employed, along with a close mockup of the power density history during self-limiting reactor transients. To date, four fuel systems have been investigated: 1) UC.ZrC particles in graphite; 2) UO 2 particles in graphite; 3) UO 2 particles in chromium 4) UO 2 particles in stainless steel. The study indicates that the maximum allowable particle diameter varies as the square root of the initial transient period and of the particle volume fraction. The critical thermophysical parameter is the thermal diffusivity of the particle, since in all cases studied it is many times smaller than that of the matrix. That of the UC.ZrC solid solution particle is 5 or more times larger than that of the UO 2 particle. It was found that the particles of system 1) above could be about 4 times larger than that of the other sy

  6. Runaway behavior and thermally safe operation of multiple liquid-liquid reactions in the semi-batch reactor. The nitric acid oxidation of 2-octanol.

    NARCIS (Netherlands)

    van Woezik, B.A.A.; Westerterp, K.R.

    2001-01-01

    The thermal runaway behavior of an exothermic, heterogeneous, multiple reaction system has been studied in a cooled semi-batch reactor. The nitric acid oxidation of 2-octanol has been used to this end. During this reaction, 2-octanone is formed, which can be further oxidized to unwanted carboxylic

  7. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  8. Molten Corium-Concrete Interaction Behavior Analyses for Severe Accident Management in CANDU Reactor

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, D. H.; Song, Y. M.

    2014-01-01

    After the last few severe accidents, the importance of accident management in nuclear power plants has increased. Many countries, including the United States (US) and Canada, have focused on understanding severe accidents in order to identify ways to further improve the safety of nuclear plants. It has been recognized that severe accident analyses of nuclear power plants will be beneficial in understanding plant-specific vulnerabilities during severe accidents. The objectives of this paper are to describe the molten corium behavior to identify a plant response with various concrete specific components. Accident analyses techniques using ISSAC can be useful tools for MCCI behavior in severe accident mitigation

  9. Investigation of mechanical properties and proton irradiation behaviors of SA-738 Gr.B steel used as reactor containment

    Directory of Open Access Journals (Sweden)

    Ma Yongzheng

    2016-08-01

    Full Text Available The proton irradiation behaviors of two kinds of SA-738Gr.B steels prepared by different heat treatment used as AP1000 reactor containment were investigated by transmission electron microscopy and positron annihilation lifetime spectrum (PAS. The mechanical properties of as-received steels were also measured. In the unirradiated conditions, the SA-738Gr.B steels had high tensile strength and excellent impact fracture toughness, which met the performance requirements of ASME codes. Both kinds of SA-738Gr.B steels were irradiated by 400keV proton from 1.07×1017H+/cm2 to 5.37×1017H+/cm2 fluence at 150 ºC. Some voids and dislocation loops with several nanometers were observed in the cross-section irradiated samples prepared by electroplating and then twin-jet electropolishing technology. The number of irradiation defects increased with increasing of displacement damage, as well as for the mean positron lifetimes. The stress-relief annealing treatment improved irradiation resistance based on open volume defect analysis from proton irradiation. SA-738Gr.B (SR steel had higher proton irradiation resistance ability than that of SA-738Gr.B (QT steel. The mechanism of irradiation behaviors were also analyzed and discussed.

  10. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

  11. POPs, Fatty acids, lipid and Stable Isotopes data - The behavioral ecology of deep-diving odontocetes in the Bahamas

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This project will use a unique set of individual-based data to quantify and model the behavioral ecology of six Department of Defense priority cetacean species in...

  12. Are tutor behaviors in problem-based learning stable?: a generalizability study of social congruence, expertise and cognitive congruence.

    Science.gov (United States)

    Williams, Judith C; Alwis, W A M; Rotgans, Jerome I

    2011-10-01

    The purpose of this study was to investigate the stability of three distinct tutor behaviors (1) use of subject-matter expertise, (2) social congruence and (3) cognitive congruence, in a problem-based learning (PBL) environment. The data comprised the input from 16,047 different students to a survey of 762 tutors administered in three consecutive semesters. Over the three semesters each tutor taught two of the same course and one different course. A generalizability study was conducted to determine whether the tutor behaviors were generalizable across the three measurement occasions. The results indicate that three semesters are sufficient to make generalizations about all three tutor behaviors. In addition the results show that individual differences between tutors account for the greatest differences in levels of expertise, social congruence and cognitive congruence. The study concludes that tutor behaviors are fairly consistent in PBL and somewhat impervious to change. Implications of these findings for tutor training are discussed.

  13. A thermodynamic assessment of the behavior of cesium and rubidium in reactor fuel elements

    International Nuclear Information System (INIS)

    Kohli, R.

    1981-01-01

    A comprehensive thermodynamic model is developed to assess the reaction and transport behavior of fission products in LWR fuel elements. The emphasis is on the chemistry of cesium and rubidium and their reactions with the fuel, other fission products, and the zircaloy cladding. Equilibrium thermodynamic calculations have been performed on the most plausible reactions to predict the chemical state of the fission products. The relevance of the predictions to pellet-clad interaction failures is discussed in detail. (orig.)

  14. Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Carbajal R, M. I.

    2012-01-01

    Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + αD + βD. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y=αD. The dose-response relationship is established in blood samples from the same donor, the coefficient α of the dose-response is Y = (0.3692 ± 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

  15. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  16. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

    Science.gov (United States)

    Le Guillou, M.; Toulhoat, N.; Pipon, Y.; Moncoffre, N.; Khodja, H.

    2015-06-01

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-cooled nuclear fission reactors (called UNGG for "Uranium Naturel-Graphite-Gaz") to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D+ ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO2) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the total amount produced

  17. The Effect of non-Newtonian Behavior of Fluid on the Tubular Reactor Performance and Studying Different Variables in Relation to the Degree of Conversion

    Directory of Open Access Journals (Sweden)

    Keivan Shayesteh

    2015-02-01

    Full Text Available In tubular reactors, there are different parameters which can affect the degree of conversion. The type of fluid motion and velocity profile of substances in the reactor are of most central measures. Different rheological models can be employed to study the behavior of fluids; power law model is one of the most commonly used models. In this study, the rheological behaviour of polymerization reaction of methyl methacrylate was examined. Due to the similarity function of tubular and batch reactors, the number of test tubes are used to prepare the solution. After preparation of the reactor solution, the n value of power law model was estimated within the span of 0.3492 to 0.9889 by curve fitting. Employing these rheological data, a reactor has been designed. Moreover, the effects of parameters such as reaction temperature, initiator wt%, the concentration of monomer and reactor’s radius on the degree of conversion have been studied. The obtained results in the research indicate a direct proportionality of conversion with the reaction temperature, initiator wt% and the concentration of monomer and also an inverse proportionality of conversion with reactor’s radius. Finally, the amount of conversion was obtained equal to 56.47% and according to its laboratory proportion which was 55.88% we have reached the conclusion that the modeling duly undertaken is applicable and valid.

  18. Hot Deformation Behavior of SA508Gr.4N Steel for Reactor Pressure Vessels

    Directory of Open Access Journals (Sweden)

    YANG Zhi-qiang

    2017-08-01

    Full Text Available The high-temperature plastic deformation and dynamic recrystallization behavior of SA508Gr.4N steel were investigated through hot deformation tests in a Gleeble1500D thermal mechanical simulator. The compression tests were performed in the temperature range of 1050-1250℃ and the strain rate range of 0.001-0.1s-1 with true strain of 0.16. The results show that from the high-temperature true stress-strain curves of the SA508Gr.4N steel, the main feature is dynamic recrystallization,and the peak stress increases with the decrease of deformation temperature or the increase of strain rate, indicating the experimental steel is temperature and strain rate sensitive material. The constitutive equation for SA508Gr.4N steel is established on the basis of the true stress-strain curves, and exhibits the characteristics of the high-temperature flow behavior quite well, while the activation energy of the steel is determined to be 383.862kJ/mol. Furthermore, an inflection point is found in the θ-σ curve, while the -dθ/dσ-σ curve shows a minimum value. The critical strain increases with increasing strain rate and decreasing deformation temperature. A linear relationship between critical strain (εc and peak strain (εp is found and could be expressed as εc/εp=0.517. The predicted model of critical strain could be described as εc=8.57×10-4Z0.148.

  19. Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.; Xiang, J.Y.

    1995-01-01

    Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment

  20. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  1. Mass-transfer behavior of materials for liquid-metal fast breeder reactor steam-generators, (1)

    International Nuclear Information System (INIS)

    Matsumoto, Keiji; Ohta, Yoshio; Kataoka, Tadayuki; Suzuki, Katsumi; Yagi, Shigeji

    1976-01-01

    To examine the compatibility of structural materials with liquid sodium in the Liquid-Metal Fast Breeder Reactor (LMFBR) steam generator environments, tests were conducted, in a sodium loop facility, on selected steels mainly of ferrite 2 1/4Cr-1Mo, 2 1/4Cr-1Mo-Nb, high Cr-1Mo, and austenitic AISI type 316. As a result of the above tests, it was found that the decarburization behavior is conditioned by the heat treatment history of steel, and the carbon loss is more rapid in annealed steel than in normalized and tempered steels. Other noteworthy findings were as follows: the carbon concentration immediately beneath the specimen surface increased consistently with the chromium content of steel, and within the range of chromium content of 3 to 5 % weight the decarburization shown by the low-chromium specimens changed to carburization. The tensile strength of annealed 2 1/4Cr-1Mo steel at 500 0 C was markedly reduced by thermal aging rather than by sodium exposure. The influence of sodium exposure on the decrease of tensile strength was slight. The creep rupture strength at 550 0 C was decreased by some 10% by sodium exposure. (auth.)

  2. An Investigation on Cocombustion Behaviors of Hydrothermally Treated Municipal Solid Waste with Coal Using a Drop-Tube Reactor

    Directory of Open Access Journals (Sweden)

    Liang Lu

    2012-01-01

    Full Text Available This work aims at demonstrating the feasibility of replacing Indonesian coal (INC with hydrothermally treated municipal solid waste (MSWH in cocombustion with high ash Indian coal (IC. The combustion efficiencies and emissions (CO, NO of MSWH, INC and their blends with IC for a series of tests performed under a range of temperatures and air conditions were tested in a drop-tube reactor (DTR. The results showed the following. The combustion efficiency of IC was increased by blending both MSWH and INC and CO emission was reduced with increasing temperature. For NO emission, the blending of MSWH led to the increase of NO concentration whereas the effects of INC depended on the temperature. The combustion behaviors of IC-MSWH blend were comparable to those of the IC-INC blend indicating it is possible for MSWH to become a good substitute for INC supporting IC combustion. Moreover, the CO emission fell while the NO emission rose with increasing excess air for IC-MSWH blend at 900°C and the highest combustion efficiency was obtained at the excess air of 1.9. The existence of moisture in the cocombustion system of IC-MSWH blend could slightly improve the combustion efficiency, reduce CO, and increase NO.

  3. Study on severe accident fuel dispersion behavior in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    Core flow blockage events have been determined to represent a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Hat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. The results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

  4. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

    2013-01-01

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  5. Kinetic and hysteretic behavior of ATP hydrolysis of the highly stable dimeric ATP synthase of Polytomella sp.

    Science.gov (United States)

    Villavicencio-Queijeiro, Alexa; Pardo, Juan Pablo; González-Halphen, Diego

    2015-06-01

    The F1FO-ATP synthase of the colorless alga Polytomella sp. exhibits a robust peripheral arm constituted by nine atypical subunits only present in chlorophycean algae. The isolated dimeric enzyme exhibits a latent ATP hydrolytic activity which can be activated by some detergents. To date, the kinetic behavior of the algal ATPase has not been studied. Here we show that while the soluble F1 sector exhibits Michaelis-Menten kinetics, the dimer exhibits a more complex behavior. The kinetic parameters (Vmax and Km) were obtained for both the F1 sector and the dimeric enzyme as isolated or activated by detergent, and this activation was also seen on the enzyme reconstituted in liposomes. Unlike other ATP synthases, the algal dimer hydrolyzes ATP on a wide range of pH and temperature. The enzyme was inhibited by oligomycin, DCCD and Mg-ADP, although oligomycin induced a peculiar inhibition pattern that can be attributed to structural differences in the algal subunit-c. The hydrolytic activity was temperature-dependent and exhibited activation energy of 4 kcal/mol. The enzyme also exhibited a hysteretic behavior with a lag phase strongly dependent on temperature but not on pH, that may be related to a possible regulatory role in vivo. Copyright © 2015 Elsevier Inc. All rights reserved.

  6. Stable isotopes

    International Nuclear Information System (INIS)

    Evans, D.K.

    1986-01-01

    Seventy-five percent of the world's stable isotope supply comes from one producer, Oak Ridge Nuclear Laboratory (ORNL) in the US. Canadian concern is that foreign needs will be met only after domestic needs, thus creating a shortage of stable isotopes in Canada. This article describes the present situation in Canada (availability and cost) of stable isotopes, the isotope enrichment techniques, and related research programs at Chalk River Nuclear Laboratories (CRNL)

  7. Phenomenology of the behavior of nuclear fuels containing plutonium in the cycles of water reactors. Development of a model on the equivalence of Plutonium

    International Nuclear Information System (INIS)

    Azzoug, D.

    1990-05-01

    In the scope of fuel recycling, in nuclear reactors with water cooling systems, a model concerning the plutonium equivalence and adapted to the thermal spectra is proposed. The physical phenomena involving the plutonium isotopes are studied. A method based on the sensitivity analysis allows the understanding of the plutonium isotope behavior. An equivalence model of plutonium for thermal spectre is established. The validity of the model for different cycle lengths and supports is proved [fr

  8. Radiation behavior of high-entropy alloys for advanced reactors. Final report

    International Nuclear Information System (INIS)

    Liaw, Peter K.; Egami, Takeshi; Zhang, Chuan; Zhang, Fan; Zhang, Yanwen

    2015-01-01

    In the first task, we have demonstrated the radiation damage and the recrystallization behaviors in multicomponent alloys through molecular-dynamics simulations. It is found that by alloying with atoms of different sizes, the atomic-level strain increases, and the propensity of the radiation-induced crystalline to amorphous transition increases as the defects cluster in the cascade body. Recrystallization of the radiation induced supercooled or glass regions show that by tuning the composition and the equilibrium temperature, the multicomponent alloys can be healed. The crystalline-amorphous-crystalline transitions predict the potential high radiation resistance in multicomponent alloys. In the second task, three types of high-entropy alloys (HEAs) were fabricated from AlCoCrFeNi and AlCuCrFeNi quinary alloys. Hardness and reduced contact modulus were measured using nanoindentation tests. Heavy ion irradiation were performed using 10 MeV gold and 5 MeV nickel to study radiation effects. Al 0.5 CrCuFeNi 2 shows phase separation upon the presence of copper. Both hardness and contact modulus exhibit the same trend as increasing the applied load, and it indicates that excessive free volume may alter the growth rate of the plastic zone. The as-cast Al 0.1 CoCrFeNi specimen undergone the hot isostatic pressing (HIP) process and steady cooling rate which mitigate the quenching effect. The swelling behavior was characterized by the atomic force microscopy (AFM), and the swelling rate is approximately 0.02% dpa. Selected area diffraction (SAD) patters show irradiation-induced amorphization throughout the ion projected range. Within the peak damage region, an amorpous ring is observed, and a mixture of amorphous/ crystalline structure at deeper depth is found. The Al 0.3 CoCrFeNi HEAs shows good radiation resistance up to 60 peak dpa. No voids or dislocations are observed. The crystal structures remain face-centered-cubic (FCC) before and after 5 MeV Ni irradiation. Higher

  9. Imperial College Reactor Centre annual report 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The report covers the following matters: research topics (reactor engineering; neutron and gamma dosimetry; nuclear physics; stable and radiotracer studies; neutron activation analysis (medicine; the environment; archaeology; geology)); personnel; publications; overseas visits; research contracts; teaching; reactor operations. (U.K.)

  10. Radiation behavior of high-entropy alloys for advanced reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Liaw, Peter K. [Univ. of Tennessee, Knoxville, TN (United States); Egami, Takeshi [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Chuan [CompuTherm, LLC, Madison, WI (United States); Zhang, Fan [CompuTherm, LLC, Madison, WI (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States)

    2015-04-30

    In the first task, we have demonstrated the radiation damage and the recrystallization behaviors in multicomponent alloys through molecular-dynamics simulations. It is found that by alloying with atoms of different sizes, the atomic-level strain increases, and the propensity of the radiation-induced crystalline to amorphous transition increases as the defects cluster in the cascade body. Recrystallization of the radiation induced supercooled or glass regions show that by tuning the composition and the equilibrium temperature, the multicomponent alloys can be healed. The crystalline-amorphous-crystalline transitions predict the potential high radiation resistance in multicomponent alloys. In the second task, three types of high-entropy alloys (HEAs) were fabricated from AlCoCrFeNi and AlCuCrFeNi quinary alloys. Hardness and reduced contact modulus were measured using nanoindentation tests. Heavy ion irradiation were performed using 10 MeV gold and 5 MeV nickel to study radiation effects. Al0.5CrCuFeNi2 shows phase separation upon the presence of copper. Both hardness and contact modulus exhibit the same trend as increasing the applied load, and it indicates that excessive free volume may alter the growth rate of the plastic zone. The as-cast Al0.1CoCrFeNi specimen undergone the hot isostatic pressing (HIP) process and steady cooling rate which mitigate the quenching effect. The swelling behavior was characterized by the atomic force microscopy (AFM), and the swelling rate is approximately 0.02% dpa. Selected area diffraction (SAD) patters show irradiation-induced amorphization throughout the ion projected range. Within the peak damage region, an amorpous ring is observed, and a mixture of amorphous/ crystalline structure at deeper depth is found. The Al0.3CoCrFeNi HEAs shows good radiation resistance up to 60 peak dpa. No voids or dislocations are observed. The crystal structures remain face-centered-cubic (FCC) before and

  11. Numerical models for the analysis of thermal behavior and coolability of a particulate debris bed in reactor lower head

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Sang Baik; Kim, Byung Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This report provides three distinctive, but closely related numerical models developed for the analysis of thermal behavior and coolability of a particulate debris bed that is may be formed inside the reactor lower head during severe accident late phases. The first numerical module presented in the report, MELTPRO-DRY, is used to analyze numerically heat-up and melting process of the dry particle bed, downward- and sideward-relocation of the liquid melt under gravity force and capillary force acting among porous particles, and solidification of the liquid melt relocated into colder region. The second module, MELTPROG-WET, is used to simulate numerically the cooling process of the particulate debris bed under the existence of water, which is subjected to two types of numerical models. The first type of WET module utilizes distinctive models that parametrically simulate the water cooling process, that is, quenching region, dryout region, and transition region. The choice of each parametric model depends on temperature gradient between the cooling water and the debris particles. The second type of WET module utilizes two-phase flow model that mechanically simulates the cooling process of the debris bed. For a consistent simulation from the water cooling to the dryout debris bed, on the other hand, the aforementioned two modules, MELTPROG-DRY and MELTPROG-WET, were integrated into a single computer program DBCOOL. Each of computational models was verified through limited applications to a heat-generating particulate bed contained in the rectangular cavity. 22 refs., 5 figs., 2 tabs. (Author)

  12. Statistic analysis of grouping in evaluation of the behavior of stable chemical elements and physical-chemical parameters in effluent from uranium mining

    International Nuclear Information System (INIS)

    Pereira, Wagner de S.

    2013-01-01

    The Ore Treatment Unit (UTM) is a uranium mine off. The statistical analysis of clustering was used to evaluate the behavior of stable chemical elements and physico-chemical variables in their effluents. The use of cluster analysis proved effective in the evaluation, allowing to identify groups of chemical elements in physico-chemical variables and group analyzes (element and variables ). As a result, we can say, based on the analysis of the data, a strong link between Ca and Mg and between Al and TR 2 O 3 (rare earth oxides) in the UTM effluents. The SO 4 was also identified as strongly linked to total solids and dissolved and these linked to electrical conductivity. Other associations existed, but were not as strongly linked. Additional collections for seasonal evaluation are required so that assessments can be confirmed. Additional statistics analysis (ordination techniques) should be used to help identify the origins of the groups identified in this analysis. (author)

  13. Adsorption Behavior of High Stable Zr-Based MOFs for the Removal of Acid Organic Dye from Water

    Directory of Open Access Journals (Sweden)

    Ke-Deng Zhang

    2017-02-01

    Full Text Available Zirconium based metal organic frameworks (Zr-MOFs have become popular in engineering studies due to their high mechanical stability, thermostability and chemical stability. In our work, by using a theoretical kinetic adsorption isotherm, we can exert MOFs to an acid dye adsorption process, experimentally exploring the adsorption of MOFs, their external behavior and internal mechanism. The results indicate their spontaneous and endothermic nature, and the maximum adsorption capacity of this material for acid orange 7 (AO7 could be up to 358 mg·g−1 at 318 K, estimated by the Langmuir isotherm model. This is ascribed to the presence of an open active metal site that significantly intensified the adsorption, by majorly increasing the interaction strength with the adsorbates. Additionally, the enhanced π delocalization and suitable pore size of UiO-66 gave rise to the highest host–guest interaction, which further improves both the adsorption capacity and separation selectivity at low concentrations. Furthermore, the stability of UiO-66 was actually verified for the first time, through comparing the structure of the samples before and after adsorption mainly by Powder X-ray diffraction and thermal gravimetric analysis.

  14. A cubic autocatalytic reaction in a continuous stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yakubu, Aisha Aliyu; Yatim, Yazariah Mohd [School of Mathematical Sciences, Universiti Sains Malaysia, 11800 USM, Penang Malaysia (Malaysia)

    2015-10-22

    In the present study, the dynamics of the cubic autocatalytic reaction model in a continuous stirred tank reactor with linear autocatalyst decay is studied. This model describes the behavior of two chemicals (reactant and autocatalyst) flowing into the tank reactor. The behavior of the model is studied analytically and numerically. The steady state solutions are obtained for two cases, i.e. with the presence of an autocatalyst and its absence in the inflow. In the case with an autocatalyst, the model has a stable steady state. While in the case without an autocatalyst, the model exhibits three steady states, where one of the steady state is stable, the second is a saddle point while the last is spiral node. The last steady state losses stability through Hopf bifurcation and the location is determined. The physical interpretations of the results are also presented.

  15. Development and verification of a modular program system for calculation of the long-time behavior of pressurized water reactors

    International Nuclear Information System (INIS)

    Woerner, A.

    1984-01-01

    For the fuel management of nuclear power plants the time dependent fuel composition and power distribution within the reactor core has to be determined with high accuracy. The calculation of this long-time behaviour of the pressurized water reactors, taking into account the thermohydraulic feedback on the above mentioned distributions, has been integrated in a modular program system. The essential part of this program system is a newly developped nodal diffusion theory program which allows to describe approximately the local variation of cross sections within a node. The above mentioned computational method, making use of latest cross section data, permits reactor cycle calculations with high accuracy. This has been demonstrated by the verification of two reactor cycles of the Kernkraftwerk Obrigheim. In addition to this physical verification a benchmark problem comparison has been accomplished to prove the efficiency of the modular program system. (orig./HP) [de

  16. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  17. Fast Pyrolysis Behavior of Banagrass as a Function of Temperature and Volatiles Residence Time in a Fluidized Bed Reactor.

    Science.gov (United States)

    Morgan, Trevor James; Turn, Scott Q; George, Anthe

    2015-01-01

    A reactor was designed and commissioned to study the fast pyrolysis behavior of banagrass as a function of temperature and volatiles residence time. Four temperatures between 400 and 600°C were examined as well as four residence times between ~1.0 and 10 seconds. Pyrolysis product distributions of bio-oil, char and permanent gases were determined at each reaction condition. The elemental composition of the bio-oils and chars was also assessed. The greatest bio-oil yield was recorded when working at 450°C with a volatiles residence time of 1.4 s, ~37 wt% relative to the dry ash free feedstock (excluding pyrolysis water). The amounts of char (organic fraction) and permanent gases under these conditions are ~4 wt% and 8 wt% respectively. The bio-oil yield stated above is for 'dry' bio-oil after rotary evaporation to remove solvent, which results in volatiles and pyrolysis water being removed from the bio-oil. The material removed during drying accounts for the remainder of the pyrolysis products. The 'dry' bio-oil produced under these conditions contains ~56 wt% carbon which is ~40 wt% of the carbon present in the feedstock. The oxygen content of the 450°C, 1.4 s 'dry' bio-oil is ~38 wt%, which accounts for ~33 wt% of the oxygen in the feedstock. At higher temperature or longer residence time less bio-oil and char is recovered and more gas and light volatiles are produced. Increasing the temperature has a more significant effect on product yields and composition than increasing the volatiles residence time. At 600°C and a volatiles residence time of 1.2 seconds the bio-oil yield is ~21 wt% of the daf feedstock, with a carbon content of 64 wt% of the bio-oil. The bio-oil yield from banagrass is significantly lower than from woody biomass or grasses such as switchgrass or miscanthus, but is similar to barley straw. The reason for the low bio-oil yield from banagrass is thought to be related to its high ash content (8.5 wt% dry basis) and high concentration of alkali

  18. Fast Pyrolysis Behavior of Banagrass as a Function of Temperature and Volatiles Residence Time in a Fluidized Bed Reactor.

    Directory of Open Access Journals (Sweden)

    Trevor James Morgan

    Full Text Available A reactor was designed and commissioned to study the fast pyrolysis behavior of banagrass as a function of temperature and volatiles residence time. Four temperatures between 400 and 600°C were examined as well as four residence times between ~1.0 and 10 seconds. Pyrolysis product distributions of bio-oil, char and permanent gases were determined at each reaction condition. The elemental composition of the bio-oils and chars was also assessed. The greatest bio-oil yield was recorded when working at 450°C with a volatiles residence time of 1.4 s, ~37 wt% relative to the dry ash free feedstock (excluding pyrolysis water. The amounts of char (organic fraction and permanent gases under these conditions are ~4 wt% and 8 wt% respectively. The bio-oil yield stated above is for 'dry' bio-oil after rotary evaporation to remove solvent, which results in volatiles and pyrolysis water being removed from the bio-oil. The material removed during drying accounts for the remainder of the pyrolysis products. The 'dry' bio-oil produced under these conditions contains ~56 wt% carbon which is ~40 wt% of the carbon present in the feedstock. The oxygen content of the 450°C, 1.4 s 'dry' bio-oil is ~38 wt%, which accounts for ~33 wt% of the oxygen in the feedstock. At higher temperature or longer residence time less bio-oil and char is recovered and more gas and light volatiles are produced. Increasing the temperature has a more significant effect on product yields and composition than increasing the volatiles residence time. At 600°C and a volatiles residence time of 1.2 seconds the bio-oil yield is ~21 wt% of the daf feedstock, with a carbon content of 64 wt% of the bio-oil. The bio-oil yield from banagrass is significantly lower than from woody biomass or grasses such as switchgrass or miscanthus, but is similar to barley straw. The reason for the low bio-oil yield from banagrass is thought to be related to its high ash content (8.5 wt% dry basis and high

  19. Fast Pyrolysis Behavior of Banagrass as a Function of Temperature and Volatiles Residence Time in a Fluidized Bed Reactor

    Science.gov (United States)

    Morgan, Trevor James; Turn, Scott Q.; George, Anthe

    2015-01-01

    A reactor was designed and commissioned to study the fast pyrolysis behavior of banagrass as a function of temperature and volatiles residence time. Four temperatures between 400 and 600°C were examined as well as four residence times between ~1.0 and 10 seconds. Pyrolysis product distributions of bio-oil, char and permanent gases were determined at each reaction condition. The elemental composition of the bio-oils and chars was also assessed. The greatest bio-oil yield was recorded when working at 450°C with a volatiles residence time of 1.4 s, ~37 wt% relative to the dry ash free feedstock (excluding pyrolysis water). The amounts of char (organic fraction) and permanent gases under these conditions are ~4 wt% and 8 wt% respectively. The bio-oil yield stated above is for 'dry' bio-oil after rotary evaporation to remove solvent, which results in volatiles and pyrolysis water being removed from the bio-oil. The material removed during drying accounts for the remainder of the pyrolysis products. The 'dry' bio-oil produced under these conditions contains ~56 wt% carbon which is ~40 wt% of the carbon present in the feedstock. The oxygen content of the 450°C, 1.4 s 'dry' bio-oil is ~38 wt%, which accounts for ~33 wt% of the oxygen in the feedstock. At higher temperature or longer residence time less bio-oil and char is recovered and more gas and light volatiles are produced. Increasing the temperature has a more significant effect on product yields and composition than increasing the volatiles residence time. At 600°C and a volatiles residence time of 1.2 seconds the bio-oil yield is ~21 wt% of the daf feedstock, with a carbon content of 64 wt% of the bio-oil. The bio-oil yield from banagrass is significantly lower than from woody biomass or grasses such as switchgrass or miscanthus, but is similar to barley straw. The reason for the low bio-oil yield from banagrass is thought to be related to its high ash content (8.5 wt% dry basis) and high concentration of alkali

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  1. The Stable Concordance Genus

    OpenAIRE

    Kearney, M. Kate

    2013-01-01

    The concordance genus of a knot is the least genus of any knot in its concordance class. Although difficult to compute, it is a useful invariant that highlights the distinction between the three-genus and four-genus. In this paper we define and discuss the stable concordance genus of a knot, which describes the behavior of the concordance genus under connected sum.

  2. A simplified analysis of granule behavior in ASBR and UASB reactors treating low-strength synthetic wastewater

    Directory of Open Access Journals (Sweden)

    R. G. Veronez

    2005-09-01

    Full Text Available This work presents an analysis of the changes observed in granule characteristics of sludge in the treatment of synthetic wastewater at a concentration of about 500 mgCOD/L in batch, fed-batch (ASBR and continuous (UASB bench-scale reactors under similar experimental conditions. Physical and microbiological properties of the granules were characterized as average particle size and sedimentation time and by optical and epifluorescence microscopy. Several samples were analyzed in order to identify the morphologies. Granules from sequencing batch and fed-batch reactors, either with or without mechanical mixing, did not undergo any physical or microbiological changes. However, during the experiment granules from the UASB reactor agglomerated due to the formation and accumulation of a viscous material, probably of microbial origin, when operated at low superficial velocities (0.072, 0.10 and 0.19 m/h. When the superficial velocity was increased to 8.0-10.0 m/h by means of liquid-phase recirculation, the granules from the UASB reactor underwent flocculation and the microbiological characteristics changed in such a way that the equilibrium of microbial diversity in the inoculum was not maintained. As a result, the only reactor that maintained efficiency and good solids retention during the assays was the ASBR, showing that there is a correlation between maintenance of microbial diversity and operating mode in the case of anaerobic treatment of low-strength wastewaters.

  3. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1978-01-01

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

  4. Nonlinear stability models and analyses of the nuclear-coupled thermal-hydraulic behavior of boiling water reactors. Interim report, October 1, 1992--May 30, 1993

    International Nuclear Information System (INIS)

    Dorning, J.J.

    1993-05-01

    All the objectives originally scheduled for the first year of this grant have been achieved. Furthermore, the project is ahead of schedule, in that a substantial amount of work has been completed on two significant objectives originally planned for the second year. This interim report is divided into five parts, summarizing the mathematical development, analysis and results of the project goals -- goals originally planned for the first year and completed, and those on which substantial progress has been made ahead of schedule. Effects of unheated riser sections and the downcomer recirculation loop on the stability characteristics of advanced boiling water reactor designs that incorporate risers or unheated channel extensions are summarized in Part A. Such extensions are incorporated above the heated reactor core channels to enhance buoyancy-driven natural thermal convection both during normal at-power operation and during emergency shutdown. The effects of both, the inclusion of unheated riser sections in the designs (one of the goals substantially completed ahead of schedule), and the inclusion of the recirculation loop in the models (first year goal) were generally found to be destabilizing. In general, as riser lengths were increased equilibria that previously were stable became unstable, and the systems with the taller risers evolved to density-wave limit cycle oscillations. As a building block of the second year goal -- to extend the one dimensional dynamical analysis of reactor thermal-hydraulics/neutron-kinetics to two and three dimensions -- we have carried out, ahead of schedule, the nonlinear dynamical analysis of two-phase flow in multiple parallel heated channels. Some basic aspects of bifurcation phenomena in two-phase flow and the related nonlinear dynamics of single and multiple parallel, uniformly and nonuniformly heated channels are studied

  5. Stable particles

    International Nuclear Information System (INIS)

    Samios, N.P.

    1994-01-01

    I have been asked to review the subject of stable particles, essentially the particles that eventually comprised the meson and baryon octets, with a few more additions - with an emphasis on the contributions made by experiments utilizing the bubble chamber technique. In this activity, much work had been done by the photographic emulsion technique and cloud chambers - exposed to cosmic rays as well as accelerator based beams. In fact, many if not most of the stable particles were found by these latter two techniques, however, the foree of the bubble chamber (coupled with the newer and more powerful accelerators) was to verify, and reinforce with large statistics, the existence of these states, to find some of the more difficult ones, mainly neutrals and further to elucidate their properties, i.e., spin, parity, lifetimes, decay parameters, etc. (orig.)

  6. Stable particles

    International Nuclear Information System (INIS)

    Samios, N.P.

    1993-01-01

    I have been asked to review the subject of stable particles, essentially the particles that eventually comprised the meson and baryon octets. with a few more additions -- with an emphasis on the contributions made by experiments utilizing the bubble chamber technique. In this activity, much work had been done by the photographic emulsion technique and cloud chambers-exposed to cosmic rays as well as accelerator based beams. In fact, many if not most of the stable particles were found by these latter two techniques, however, the forte of the bubble chamber (coupled with the newer and more powerful accelerators) was to verify, and reinforce with large statistics, the existence of these states, to find some of the more difficult ones, mainly neutrals and further to elucidate their properties, i.e., spin, parity, lifetimes, decay parameters, etc

  7. H Reactor

    Data.gov (United States)

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

  8. The effect of reactor vessel vent valves (RVVV's) on SBLOCA behavior of the UMCP 2 x 4 B and W simulation loop

    International Nuclear Information System (INIS)

    Pertmer, G.A.; Almenas, K.; DiMarzo, M.; Bailery, S.

    1989-01-01

    This paper presents results of a series of Small Break LOCA tests performed at the University of Maryland 2 x 4 Loop, which is a 1/500 volume scaled model of a B and W PWR system. A feature of the B and W system which has a major impact on SBLOCA behavior is the Reactor Vessel Vent Valves, which provide a flow path between the upper plenum and downcomer in the vessel. This paper compares system behavior during a SBLOCA for two tests - one in which the valves are allowed to open, and one in which the valves are held closed. Similarities and differences between the system flow modes that occur during the transient are presented and discussed for the two tests

  9. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm2)

    International Nuclear Information System (INIS)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J.

    1999-01-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm 2 ). (Author)

  10. THE BEHAVIOR OF SOLUBLE METALS ELUTED FROM Ni/Fe-BASED ALLOY REACTORS AFTER HIGH-TEMPERATURE AND HIGH-PRESSURE WATER PROCESS

    Directory of Open Access Journals (Sweden)

    M. Faisal

    2012-05-01

    Full Text Available The behavior of heavy metals eluted from the wall of Ni/Fe-based alloy reactors after high-temperature and high-pressure water reaction were studied at temperatures ranging from 250 to 400oC. For this purpose, water and cysteic acid were heated in two reactor materials which are SUS 316 and Inconel 625. Under the tested conditions, the erratic behaviors of soluble metals eluted from the wall of Ni/Fe-based alloy in high temperature water were observed. Results showed that metals could be eluted even at a short contact time. The presence of air also promotes elution at sub-critical conditions. At sub-critical conditions, a significant amount of Cr was extracted from SUS 316, while only traces of Ni, Fe, Mo and Mn were eluted. In contrast, Ni was removed in significant amounts compared to Cr when Inconel 625 was tested. It was observed that eluted metals tend to increased under acidic conditions and most of those metals were over the limit of WHO guideline for drinking water. The results are significant both on the viewpoint of environmental regulation on disposal of wastes containing heavy metals, toxicity of resulting product and catalytic effect on a particular reaction.

  11. Stable beams

    CERN Multimedia

    2015-01-01

    Stable beams: two simple words that carry so much meaning at CERN. When LHC page one switched from "squeeze" to "stable beams" at 10.40 a.m. on Wednesday, 3 June, it triggered scenes of jubilation in control rooms around the CERN sites, as the LHC experiments started to record physics data for the first time in 27 months. This is what CERN is here for, and it’s great to be back in business after such a long period of preparation for the next stage in the LHC adventure.   I’ve said it before, but I’ll say it again. This was a great achievement, and testimony to the hard and dedicated work of so many people in the global CERN community. I could start to list the teams that have contributed, but that would be a mistake. Instead, I’d simply like to say that an achievement as impressive as running the LHC – a machine of superlatives in every respect – takes the combined effort and enthusiasm of everyone ...

  12. IAEA coordinated research programme on heat transfer behavior and thermo-hydraulics code testing for super critical water cooled reactors

    International Nuclear Information System (INIS)

    Bilbao y Leon, Sama; Aksan, Nusret

    2009-01-01

    One of the key roles of the IAEA is to foster the collaboration among Member States on the development of advances in technology for advanced nuclear power plants. There is high international interest, both in developing and industrialized countries, in innovative supercritical water-cooled reactors (SCWRs), primarily because such concepts will achieve high thermal efficiencies (44-45%) and promise improved economic competitiveness utilizing and building upon the recent developments for highly efficient fossil power plants. The SCWR has been selected as one of the promising concepts for development by the Generation-IV International Forum. Following the advice of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA has recently started a Coordinated Research Programme (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The first Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna, Austria in July 2008. This paper summarizes the current status of the CRP, including the Integrated Research Plan and the general schedule for the CRP. (author)

  13. Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Wehner, T.R.

    1980-01-01

    The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

  14. Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

    International Nuclear Information System (INIS)

    Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak; Park, Young Soo

    2010-01-01

    The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

  15. Development of Mathematical Model and Analysis Code for Estimating Drop Behavior of the Control Rod Assembly in the Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Oh, Se-Hong; Kang, SeungHoon; Choi, Choengryul; Yoon, Kyung Ho; Cheon, Jin Sik

    2016-01-01

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. There are three typical ways to estimate the drop behavior of the control rod assembly in scram action: Experimental, numerical and theoretical methods. But experimental and numerical(CFD) method require a lot of cost and time. Thus, these methods are difficult to apply to the initial design process. In this study, mathematical model and theoretical analysis code have been developed in order to estimate drop behavior of the control rod assembly to provide the underlying data for the design optimization. Mathematical model and theoretical analysis code have been developed in order to estimate drop behavior of the control rod assembly to provide the underlying data for the design optimization. A simplified control rod assembly model is considered to minimize the uncertainty in the development process. And the hydraulic circuit analysis technique is adopted to evaluate the internal/external flow distribution of the control rod assembly. Finally, the theoretical analysis code(named as HEXCON) has been developed based on the mathematical model. To verify the reliability of the developed code, CFD analysis has been conducted. And a calculation using the developed analysis code was carried out under the same condition, and both results were compared

  16. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1979-01-01

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

  17. Membrane assisted fluidized bed reactors: Potentials and hurdles

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Heinrich, S.; Mörl, L.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    Recent advances in the development of more stable membranes with increased permeance have significantly enhanced the possibilities for integrating membranes into catalytic reactors in order to achieve a major increase in reactor performance by process integration and process intensification. Several

  18. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  19. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  1. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  2. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  3. Study on uranium metallization yield of spent Pressurized Water Reactor fuels and oxidation behavior of fission products in uranium metals

    International Nuclear Information System (INIS)

    Choi, Ke Chon; Lee, Chang Heon; Kim, Won Ho

    2003-01-01

    Metallization yield of uranium oxide to uranium metal from lithium reduction process of spent Pressurized Water Reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metallization yield was measured. Metallization yield of the solid part was 90.7∼95.9 wt%, and the powder being 77.8∼71.5 wt% individually. Oxidation behaviour of the quarternary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At 600∼700 .deg. C, weight increments of allow of No, Ru, Rh and Pd was 0.40∼0.55 wt%. Phase change on the surface of the allow was started at 750 .deg. C. In particular, Mo was rapidly oxidized and then the alloy lost 0.76∼25.22 wt% in weight

  4. Corrosion Behavior of Carbon Steel Coated with Octadecylamine in the Secondary Circuit of a Pressurized Water Reactor

    Science.gov (United States)

    Jäppinen, Essi; Ikäläinen, Tiina; Järvimäki, Sari; Saario, Timo; Sipilä, Konsta; Bojinov, Martin

    2017-12-01

    Corrosion and particle deposition in the secondary circuits of pressurized water reactors can be mitigated by alternative water chemistries featuring film-forming amines. In the present work, the corrosion of carbon steel in secondary side water with or without octadecylamine (ODA) is studied by in situ electrochemical impedance spectroscopy, combined with weight loss/gain measurements, scanning electron microscopy and glow-discharge optical emission spectroscopy. The impedance spectra are interpreted using the mixed-conduction model to extract kinetic parameters of oxide growth and metal dissolution through it. From the experimental results, it can be concluded that ODA addition reduces the corrosion rate of both fresh and pre-oxidized carbon steel in secondary circuit significantly by slowing down both interfacial reactions and transport through the oxide layer.

  5. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  6. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Bernard A. Toseland

    2000-12-31

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large-diameter reactors. Washington University's work during the reporting period involved the implementation of the automated calibration device, which will provide an advanced method of determining liquid and slurry velocities at high pressures. This new calibration device is intended to replace the original calibration setup, which depended on fishing lines and hooks to position the radioactive particle. The report submitted by Washington University contains a complete description of the new calibration device and its operation. Improvements to the calibration program are also discussed. Iowa State University utilized air-water bubble column simulations in an effort to determine the domain size needed to represent all of the flow scales in a gas-liquid column at a high superficial velocity. Ohio State's report summarizes conclusions drawn from the completion of gas injection phenomena studies, specifically with respect to the characteristics of bubbling-jetting at submerged single orifices in liquid-solid suspensions.

  7. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

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

  8. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  9. Dynamics of TRIGA-3 Salazar Reactor

    International Nuclear Information System (INIS)

    Gallardo S, L.F.

    1990-01-01

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

  10. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  11. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  12. Reactor vessel

    NARCIS (Netherlands)

    Makkee, M.; Kapteijn, F.; Moulijn, J.A.

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  13. Fluidization behavior in a circulating slugging fluidized bed reactor. Part I: Residence time and residence time distribution of polyethylene solids

    NARCIS (Netherlands)

    van Putten, I.C.; van Sint Annaland, M.; Weickert, G.

    2007-01-01

    Square nosed slugging fluidization behavior in a circulating fluidized bed riser using a polyethylene powder with a very wide particle size distribution was studied. In square nosed slugging fluidization the extent of mixing of particles of different size depends on the riser diameter, gas velocity,

  14. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  16. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  17. 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)

  18. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Nagatomi, Shozo.

    1976-01-01

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

  19. 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

  20. Radioisotope production in fusion reactors

    International Nuclear Information System (INIS)

    Engholm, B.A.; Cheng, E.T.; Schultz, K.R.

    1986-01-01

    Radioisotope production in fusion reactors is being investigated as part of the Fusion Applications and Market Evaluation (FAME) study. /sup 60/Co is the most promising such product identified to date, since the /sup 60/Co demand for medical and food sterilization is strong and the potential output from a fusion reactor is high. Some of the other radioisotopes considered are /sup 99/Tc, /sup 131/l, several Eu isotopes, and /sup 210/Po. Among the stable isotopes of interest are /sup 197/Au, /sup 103/Rh and Os. In all cases, heat or electricity can be co-produced from the fusion reactor, with overall attractive economics

  1. Study of REE behaviors, fluid inclusions, and O, S stable Isotopes in Zafar-abad iron skarn deposit, NW Divandarreh, Kordestan province

    Directory of Open Access Journals (Sweden)

    Mehrdad Barati

    2014-10-01

    Full Text Available Introduction The Zafar-abad iron ore deposit, situated in the NW part of Divandarreh (lat. 36°01'14" and long. 46°58'22". The ore body is located on the northern margin of the Sanandaj-Sirjan igneous metamorphic zone. The Zafar-abad Fe-skarn deposit is one of the important, medium- size mineral deposits in western Iran. REE patterns of skarn magnetite were among others studied in Skarn deposit by (Taylor, 1979 Hydrothermal alteration and fluid-rock interaction significantly affect total contents of REE and their patterns in fluids. Moreover, fractionation of REE by chemical complication, adsorption effects and redox reactions are characteristic processes determining REE behavior during crystallization. Stable isotope data for oxygen and sulfur have been widely used with great success to trace the origin and evolution history of paleo-hydrothermal fluids of meteoric, magmatic, and metamorphic. Materials and methods The present study investigates REE and stable Isotope geochemistry of magnetite and pyrite in Zafar-abad deposit and temperature of trapped fluid inclusions based on geothermometry analysis. In order to study the major, trace and REE compositions of Zafar-abad magnetite, twelve samples were collected from surface of ore exposures. The emphasis during sampling was on ores with primary textures. Discussion The Zafar-abad district is situated in Mesozoic and Cenozoic sedimentary, meta-sedimentary and meta-igneous rocks in Sanandaj-Sirjan igneous metamorphic zone. Sedimentary sequences dominantly composed of calcareous and conglomerate rocks. Various meta-sedimentary rocks are intercalated with the sedimentary rocks, and comprise biotite and muscovite-rich schist, calc-schist, calc-silicate rock. Several distinct ductile tectonic fabrics have been identified around the Zafar-abad deposit. The main ore body at Zafar-abad is in the form of a roughly horizontal, discordant, lens to tabular-shaped body plunging 10° NW, where it appears to

  2. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

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

  3. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    1980-06-01

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

  4. Fusion reactor materials

    International Nuclear Information System (INIS)

    Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

    1980-01-01

    Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

  5. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

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

  6. Halden reactor project

    International Nuclear Information System (INIS)

    1980-01-01

    The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

  7. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    2001-01-01

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

  8. Some trends in constitutive equation model development for high-temperature behavior of fast-reactor structural alloys

    International Nuclear Information System (INIS)

    Pugh, C.E.; Robinson, D.N.

    1977-01-01

    The paper addresses some important features of the inelastic behavior of 2 1 / 4 Cr--1Mo steel and indicates a mathematical framework that is capable of representing these types of response. While the constitutive model discussed embraces capabilities beyond those of equations presently used in design analyses; their implementation into practicable analysis methods (such as finite-element programs) is more demanding. For example, in the case of slow time-dependent deformations, the equations governing accumulation of the inelastic strain components and the evolution of the tensorial state variable α are intimately coupled. A part of recommending any such model for use in design must be a quantitative assessment of the economic feasibility of implementation

  9. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  10. Chaotic behavior in a system simulating the pressure balanced injection system. Analysis of passive safety reactor behavior. JAERI's nuclear research promotion program, H12-012 (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Madarame, Haruki; Okamoto, Koji; Tanaka, Gentaro; Morimoto, Yuichiro [Tokyo Univ., School of Engineering, Tokyo (Japan); Sato, Akira [Yamagata Univ., Faculty of Engineering, Yonezawa, Yamagata (Japan); Kondou, Masaya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The pressure Balanced Injection System (PBIS) was proposed in a passive safety reactor. Pressurizing Line (PL) connects the Reactor Vessel (RV) and the gas area in the Contain Vessel (CV), and Injected Line (IL) connects two vessels at relatively lower position. In an accident, the two lines are passively opened. The vapor generated by the residual heat pressed downward the water level in the RV. When the level is lower than the inlet of the PL, vapor is ejected into the CV through the PL attaining the pressure balance between the vessels. Then boron water in the CV is injected into the RV through the IL by the static head. This process is repeated by the succeeding vapor generation. In an experiment, the oscillating system was replaced by water column in a U-shaped duct. The vapor generation was simulated by cover gas supply to one end of the duct, while the other end was open to the atmosphere. When the water level reached a certain level, electromagnetic valves opened and the cover gas was ejected. The gas pressure decreased rapidly, resulting in a surface rise. When the water level reached another level, the valves closed. The cover gas pressure increased again, thus, gas ejection occurred intermittently. The interval of the gas ejection was not constant but fluctuated widely. Mere stochastic noise could hardly explain the large amplitude. Then was expressed the system using a set of linear equations. Various types of piecewise linear model were developed to examine the cause of the fluctuation. There appeared tangential bifurcation, period-doubling bifurcation, period-adding bifurcation and so on. The calculated interval exhibited chaotic features. Thus the cause of the fluctuation can be attributed to chaotic features of the system having switching. Since the piecewise linear model was highly simplified the behavior, a quantitative comparison between the calculation and the experiment was difficult. Therefore, numerical simulation code considering nonlinear

  11. Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR; Estudio del comportamiento neutronico de un ensamble combustible con gadolinia de un reactor HPLWR

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    This work presents a neutronic study of a square assembly design of double line of fuel rods, with moderator box to center of the arrangement, for the nuclear reactor cooled with supercritical water, High Performance Light Water Reactor (HPLWR). For the fuel analyses of the reactor HPLWR the neutronic code Helios-2 was used, settling down as the first study on fuel under conditions of supercritical water that has been simulated with this code. The analyzed variables, essentials in the neutronic design of any reactor, were the infinite neutrons multiplication factor (k{infinity}) and the maximum power peaking factor (PPF{sub max}), as well as the reactivity coefficients by the fuel temperature. The k{infinity} and PPF{sub max} values were obtained under conditions in cold (293.6 K) and in hot (to 880.8 K). The tests were realized for a reference fuel assembly design, with 40 fuel rods with enrichments of 4 and 5% of U-235, and considering different concentrations of consumable poison (gadolinium - Gd{sub 2O3}) in some rods of the same assembly. The obtained results show values k{infinity} and PPF{sub max} minors to the present in the conventional light water reactors. Moreover, the reactivity coefficients by fuel temperature were verified with the purpose of satisfying the safety conditions required in the nuclear reactors. (Author)

  12. Comparison of high temperature steam oxidation behavior of Zircaloy-4 versus austenitic and ferritic steels under light water reactor safety aspects

    International Nuclear Information System (INIS)

    Leistikow, S.; Schanz, G.; Zurek, Z.

    1985-12-01

    A comparative study of the oxidation behavior of Zy-4 versus steel No. 1.4914 and steel No. 1.4970 was performed in high temperature steam. Reactor typical tube sections of all three materials were exposed on both sides to superheated steam at temperatures ranging from 600 to 1300 0 C for up to 6 h. The specimens were evaluated by gravimetry, metallography, and other methods. The results are presented in terms of weight gain, corresponding metal (wall) penetration and consumption as function of time and temperature. Concerning the corrosion resistance the ranking position of Zy-4 was between the austenitic and the ferritic steel. Because of the chosen wall dimensions Zy-4 and the austenitic steel behaved similarly in that the faster oxidation of the thicker Zy-4 cladding consumed the total wall thickness in a time equivalent to the slower oxidation of the thinner austenitic steel cladding. The ferritic steel cladding however was faster consumed because of the lower oxidation resistance and the thinner wall thickness compared to the austenitic steel. So besides oxide scale formation, oxygen diffusion into the bulk of the metal forming various oxygen-containing phases were evaluated - also in respect to their influence on mechanical cladding properties and the dimensional changes. (orig./HP) [de

  13. Pyrolysis of oil-plant wastes in a TGA and a fixed-bed reactor: Thermochemical behaviors, kinetics, and products characterization.

    Science.gov (United States)

    Chen, Jianbiao; Fan, Xiaotian; Jiang, Bo; Mu, Lin; Yao, Pikai; Yin, Hongchao; Song, Xigeng

    2015-09-01

    Pyrolysis characteristics of four distinct oil-plant wastes were investigated using TGA and fixed-bed reactor coupled with GC. TGA experiments showed that the pyrolysis behaviors were related to biomass species and heating rates. As the heating rate increased, TG and DTG curves shifted to the higher temperatures, and the comprehensive devolatilization index obviously increased. The remaining chars from TGA experiments were higher than those obtained from the fixed-bed experiments. The crack of tars at high temperatures enhanced the formation of non-condensable gases. During the pyrolysis, C-O and CO2 were the major gases. Chars FTIR showed that the functional groups of O-H, C-H(n), C=O, C-O, and C-C gradually disappeared from 400 °C on. The kinetic parameters were calculated by Coats-Redfern approach. The results manifested that the most appropriate pyrolysis mechanisms were the order reaction models. The existence of kinetic compensation effect was evident. Crown Copyright © 2015. Published by Elsevier Ltd. All rights reserved.

  14. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  15. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  16. The effects of thermal aging on material behavior and strength of CF8M in nuclear reactor coolant system

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Jae Do; Lee, Yong Seon; Park, Jung Cheol; In, Jae Hyeon; Woo, Seung Wan; Pae, Yong Tak; Nam, Uk Hui; Park, Yun Won [Yeungnam Univ., Gyeongsan (Korea, Republic of)

    1999-03-15

    The following investigations are performed in order to estimate the mechanism of the structural integrity, and the life prediction. The CF8M is observed a brittle behavior in the range of 475 .deg. C. The five classes of the thermally aged CF8M specimen are prepared using an artificially accelerated aging method. Namely, after the specimen are held for 100, 300, 900, 1800 and 3600 hrs. at 430 .deg. C respectively, the specimen are water cooled to room temperature. In addition to the thermally aged specimens the specimens associated with {delta}-phase degradation are prepared. After the specimens are maintained for 20 min, 5, 15, 50 and 150 hrs. at 700 .deg. C, respectively. which is in the range of {delta}-phase degradation, all specimens are cooled in water. The impact energy variations are measured for both the aged and virgin specimen at -173, -70, -32, 27 and 100 .deg. C, respectively, through the Charpy impact tests in addition to the hardness tests. The characteristics of the fatigue crack growth and low cycle fatigue tests are investigated using both aged and virgin specimens. Also fractured surfaces of the specimen are observed using the scanning electronic microscopy. J-R curve and J{sub IC} of the aged and virgin specimens are found J{sub IC} in order to predict the critical flaw size and fatigue life.

  17. Estimation of in-plant source term release behaviors from Fukushima daiichi reactor cores by forward method and comparison with reverse method

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Won; Rhee, Bo Wook; Song, Jin Ho; Kim, Sung Il; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012–018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3

  18. Thermally sensitive block copolymer particles prepared via aerosol flow reactor method: Morphological characterization and behavior in water.

    Science.gov (United States)

    Nykänen, Antti; Rahikkala, Antti; Hirvonen, Sami-Pekka; Aseyev, Vladimir; Tenhu, Heikki; Mezzenga, Raffaele; Raula, Janne; Kauppinen, Esko; Ruokolainen, Janne

    2012-10-23

    This work describes properties of thermo-sensitive submicron sized particles having the same chemical composition but different morphologies. These particles have been prepared with an aerosol technique using dimethylformamide solutions of linear polystyrene-block-poly(N-isopropylacrylamide-block-polystyrene, PS-b-PNIPAM-b-PS. The particles were characterized by cryo-electron microscopy, microcalorimetry, and light scattering. Block-copolymers self-assembled within the particles forming onion-like, gyroid-like, and spherical morphologies having poly(N-isopropylacrylamide) matrix and physically cross-linking polystyrene domains. The particles were dispersed in aqueous media and their behavior in water was studied both below and above the lower critical solution temperature of poly(N-isopropylacrylamide). We found out that the particles with spherical and gyroid-like morphologies swell considerably in water at 20 °C, whereas at 40 °C the particles resemble more of those studied without water treatment. Light scattering experiments showed that the particles gradually aggregate and precipitate with time at 40 °C. Microcalorimetric studies revealed for all three studied morphologies that PNIPAM undergoes a two-step transition due to the different hydration levels of PNIPAM inside and outside the particles. Thicknesses of the PS and PNIPAM layers within the onion-like particles were analyzed using the TEM micrographs by fitting a model of electron density to the integrated electron intensity data. The surface layer of the particles was found out to be PNIPAM, which was supported by light scattering and microcalorimetry. It was also found out from the TEM micrograph analysis that the width of the outmost PS layer is considerably thinner than the one in the dry state prior to immersion in water, and a degradation scheme is proposed to explain these results.

  19. Stable Isotope Data

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Tissue samples (skin, bone, blood, muscle) are analyzed for stable carbon, stable nitrogen, and stable sulfur analysis. Many samples are used in their entirety for...

  20. Variation in the large-scale organization of gene expression levels in the hippocampus relates to stable epigenetic variability in behavior.

    Directory of Open Access Journals (Sweden)

    Mark D Alter

    2008-10-01

    Full Text Available Despite sharing the same genes, identical twins demonstrate substantial variability in behavioral traits and in their risk for disease. Epigenetic factors-DNA and chromatin modifications that affect levels of gene expression without affecting the DNA sequence-are thought to be important in establishing this variability. Epigenetically-mediated differences in the levels of gene expression that are associated with individual variability traditionally are thought to occur only in a gene-specific manner. We challenge this idea by exploring the large-scale organizational patterns of gene expression in an epigenetic model of behavioral variability.To study the effects of epigenetic influences on behavioral variability, we examine gene expression in genetically identical mice. Using a novel approach to microarray analysis, we show that variability in the large-scale organization of gene expression levels, rather than differences in the expression levels of specific genes, is associated with individual differences in behavior. Specifically, increased activity in the open field is associated with increased variance of log-transformed measures of gene expression in the hippocampus, a brain region involved in open field activity. Early life experience that increases adult activity in the open field also similarly modifies the variance of gene expression levels. The same association of the variance of gene expression levels with behavioral variability is found with levels of gene expression in the hippocampus of genetically heterogeneous outbred populations of mice, suggesting that variation in the large-scale organization of gene expression levels may also be relevant to phenotypic differences in outbred populations such as humans. We find that the increased variance in gene expression levels is attributable to an increasing separation of several large, log-normally distributed families of gene expression levels. We also show that the presence of these

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  2. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    reactor was found to be stable and self-regulating. 4.2 Molten-salt reactor experiment (MSRE) [6]. As ARE had .... Continuous monitoring of the atmo- sphere was carried out by Health Physics Group of ... (4) Breeding ratio should be such that a self-sustaining operation of the reactor is achieved after accounting for losses in ...

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  4. Short communication: Relationship between serum cortisol concentration and defensive behavioral responses of dairy cows exposed to natural infestation by stable fly, Stomoxys calcitrans.

    Science.gov (United States)

    Vitela-Mendoza, I; Cruz-Vázquez, C; Solano-Vergara, J; Orihuela-Trujillo, A

    2016-12-01

    The aim of this study was conducted to evaluate the effect of natural infestation by Stomoxys calcitrans on the behavioral and adrenocortical responses of dairy cattle. Twenty Holstein cows randomly selected were individually sprayed with insecticide once every 7d, whereas no insecticide was applied to the other 20 animals. The average number of flies per cow was estimated daily, and the frequency of fly-avoidance behaviors was measured daily; plasma cortisol concentration was measured each morning. No flies were ever counted on the treated cows at any time during the experiment, whereas an average of 17.13±1.14 (±standard error) flies/d were recorded on untreated cows. Tail movement was the most frequent behavior displayed, with stamps or kicks showing the highest increment rate (41.2×) when fly population increased from zero to greater than 51 flies/cow. Cortisol concentration increased to a maximum of 56.81±39.53ng/mL with 26 to 30 flies/cow per day. Coefficients of determination between the number of flies, cortisol concentration, tail movements, and stamps or kicks were 0.73, 0.78, and 0.81, respectively. The multiple correlation coefficient was 0.90, with 81% of the variation in cortisol concentration explainable by variation in the number of flies per cow and the frequency of fly-avoidance behaviors. It was concluded that plasma cortisol concentration is linearly related to a combination of the number of flies and the frequency of fly-dislodging behaviors, producing a maximum response before reaching maximum fly loads. Copyright © 2016 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

  5. Sensitivity Analysis of Reactor Regulating System for SMART

    International Nuclear Information System (INIS)

    Jeon, Yu Lim; Kang, Han Ok; Lee, Seong Wook; Park, Cheon Tae

    2009-01-01

    The integral reactor technology is one of the Small and Medium sized Reactor (SMR) which has recently come into a spotlight due to its suitability for various fields. SMART (System integrated Modular Advanced ReacTor), a small sized integral type PWR with a rated thermal power of 330MWt is one of the advanced SMR. SMART developed by the Korea Atomic Energy Research Institute (KAERI), has a capacity to provide 40,000 m3 per day of potable water and 90 MW of electricity (Chang et al., 2000). Figure 1 shows the SMART which adopts a sensible mixture of new innovative design features and proven technologies aimed at achieving highly enhanced safety and improved economics. Design features contributing to a safety enhancement are basically inherent safety improving features and passive safety features. Fundamental thermal-hydraulic experiments were carried out during the design concepts development to assure the fundamental behavior of major concepts of the SMART systems. A TASS/SMR is a suitable code for accident and performance analyses of SMART. In this paper, we proposed a new power control logic for stable operating outputs of Reactor Regulating System (RRS) of SMART. We analyzed the sensitivity of operating parameter for various operating conditions

  6. A cholinergic-regulated circuit coordinates the maintenance and bi-stable states of a sensory-motor behavior during Caenorhabditis elegans male copulation.

    Directory of Open Access Journals (Sweden)

    Yishi Liu

    2011-03-01

    Full Text Available Penetration of a male copulatory organ into a suitable mate is a conserved and necessary behavioral step for most terrestrial matings; however, the detailed molecular and cellular mechanisms for this distinct social interaction have not been elucidated in any animal. During mating, the Caenorhabditis elegans male cloaca is maintained over the hermaphrodite's vulva as he attempts to insert his copulatory spicules. Rhythmic spicule thrusts cease when insertion is sensed. Circuit components consisting of sensory/motor neurons and sex muscles for these steps have been previously identified, but it was unclear how their outputs are integrated to generate a coordinated behavior pattern. Here, we show that cholinergic signaling between the cloacal sensory/motor neurons and the posterior sex muscles sustains genital contact between the sexes. Simultaneously, via gap junctions, signaling from these muscles is transmitted to the spicule muscles, thus coupling repeated spicule thrusts with vulval contact. To transit from rhythmic to sustained muscle contraction during penetration, the SPC sensory-motor neurons integrate the signal of spicule's position in the vulva with inputs from the hook and cloacal sensilla. The UNC-103 K(+ channel maintains a high excitability threshold in the circuit, so that sustained spicule muscle contraction is not stimulated by fewer inputs. We demonstrate that coordination of sensory inputs and motor outputs used to initiate, maintain, self-monitor, and complete an innate behavior is accomplished via the coupling of a few circuit components.

  7. Characterization of the gas releasing behaviors of catalytic pyrolysis of rice husk using potassium over a micro-fluidized bed reactor

    International Nuclear Information System (INIS)

    Liu, Yuan; Wang, Yan; Guo, Feiqiang; Li, Xiaolei; Li, Tiantao; Guo, Chenglong; Chang, Jiafu

    2017-01-01

    Highlights: • Releasing propensity of CO, CO 2 , CH 4 and H 2 was studied in a micro-fluidized bed. • Gas releasing pattern was influenced by temperature and potassium concentration. • Variations in gas forming E a are indicative of catalytic performance of potassium. - Abstract: Influence of potassium on the gas releasing behaviors during rice husk high-temperature pyrolysis was investigated under isothermal conditions in a two stage micro-fluidized bed reactor. Reaction kinetics for generating H 2 , CO, CO 2 and CH 4 was investigated based on the Friedman and model-fitting approaches. Results indicated that different gas species had different times to start and end the gas release process, particularly at 600 °C, representing different chemical routes and mechanics for generating these gas components. The resulting apparent activation energies for H 2 , CO, and CO 2 decreased from 23.10 to 12.00 kJ/mol, 15.48 to 14.03 kJ/mol and 10.14 to 7.61 kJ/mol respectively with an increase in potassium concentration from 0 to 0.5 mol/kg, while that for CH 4 increased from 16.85 to 19.40 kJ/mol. The results indicated that the addition of potassium could promote the generation reactions of H 2 , CO and CO 2 while hinder the generation reactions of CH 4 . The pyrolysis reaction was further found to be subject to the three-dimensional diffusion model for all the samples.

  8. Reactor control rod supporting structure

    International Nuclear Information System (INIS)

    Akimoto, Tokuzo; Miyata, Hiroshi.

    1984-01-01

    Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

  9. SLOWPOKE reactor

    International Nuclear Information System (INIS)

    Evans, D.J.R.; Downs, W.E.

    1974-01-01

    The SLOWPOKE reactor is described, which is a small pool type with thermal neutron fluxes ranging from 10 11 -10 12 n cm -2 sec -1 . It differs in many ways from conventional pool type, namely small critical mass, beryllium reflector and a closed reactor container. The reactor is designed as small and simply as possible, and consistently with safety and good operating practice. Access to the present model is via pneumatic irradiation tubes only, which limits the use of the facility to activation analysis, tracer production and training. (Mori, K.)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  11. Effect of Primary Cooling Water on the Hot Water Layer of a Reactor Pool

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Chae, Hee Taek; Jo, Dea Sung; Lee, Byung Chul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Many research reactors, including HANARO, have a hot water layer to reduce the radioactivity level in a pool top area. The hot water layer can keep down the ascending of radio-active matters generated nearby the reactor by the neutron irradiation. The hot water layer is a stratified water layer about 5 {approx} 10 .deg. C hotter than the lower pool water. The flow in the reactor pool become fierce, the hot water layer may be broken or become thinner due to vigorous mixing between the hot water layer and the pool water. Large amount of cooling water directly dumped into the reactor pool makes the stable water pool move violently that can have a serious effect on the hot water layer. Thus, the preliminary investigation is required to figure out the mass flow dump effect on the hot water layer. The reactor pool is so gigantic that it is hard to conduct this study by an experimental method, whereas CFD method is relatively easy to simulate even such a very large structure. In this paper when a mass flow of cooling water is dumped into the reactor pool, flow behaviors of pool water are studied by CFD method

  12. Stable convergence and stable limit theorems

    CERN Document Server

    Häusler, Erich

    2015-01-01

    The authors present a concise but complete exposition of the mathematical theory of stable convergence and give various applications in different areas of probability theory and mathematical statistics to illustrate the usefulness of this concept. Stable convergence holds in many limit theorems of probability theory and statistics – such as the classical central limit theorem – which are usually formulated in terms of convergence in distribution. Originated by Alfred Rényi, the notion of stable convergence is stronger than the classical weak convergence of probability measures. A variety of methods is described which can be used to establish this stronger stable convergence in many limit theorems which were originally formulated only in terms of weak convergence. Naturally, these stronger limit theorems have new and stronger consequences which should not be missed by neglecting the notion of stable convergence. The presentation will be accessible to researchers and advanced students at the master's level...

  13. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  14. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  15. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  16. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

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

  17. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

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

  18. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  19. The mechanics in the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.

    1998-01-01

    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  20. Propulsion reactors

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A nuclear reactor equips the recently constructed French aircraft- carrier Charles-De-Gaulle, in a few months the second nuclear submarine (SNLE) of new generation will be operational. In last october the government launched the program Barracuda which consists of 6 submarines (SNA) whose series head will be operational in 2010. The main asset of nuclear propulsion is to allow an almost unlimited autonomy: soft water, air are produced inside the submarine and the maximum time spent underwater is only limited by human capacity to cope with confinement. CEA has 3 missions concerning country defence. First the designing, the fabrication and the maintenance of weapons, secondly the supplying of fissile materials and thirdly the nuclear propulsion. A new generation of propulsion reactors is being studied and a ground installation involving a test reactor equivalent to that on board is being built. This test reactor (RES) will simulate any type of on-board reactors by adjusting temperature, pressure, flowrate and even equipment such as steam generator. This reactor will validate the technological choices for the Barracuda program. (A.C.)

  1. Behaviorism

    Science.gov (United States)

    Moore, J.

    2011-01-01

    Early forms of psychology assumed that mental life was the appropriate subject matter for psychology, and introspection was an appropriate method to engage that subject matter. In 1913, John B. Watson proposed an alternative: classical S-R behaviorism. According to Watson, behavior was a subject matter in its own right, to be studied by the…

  2. Behavior

    NARCIS (Netherlands)

    Bouwman, L.I.

    2014-01-01

    Health behaviors are people’s actions, some purposefully deployed to promote or protect health; some thoughtlessly undertaken without concern for their potential risk to health; some consciously, even defiantly, deployed regardless of consequences to health. Risk behaviors are specific forms of

  3. Behavior of radon, chemical compounds and stable elements in underground water; Comportamiento de radon, compuestos quimicos y elementos estables en agua subterranea

    Energy Technology Data Exchange (ETDEWEB)

    Lopez R, N.; Segovia, N.; Lopez, M.B.E.; Pena, P. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico); Armienta, M.A.; Godinez, L. [IGFUNAM, Ciudad Universitaria, 04510 Mexico D.F. (Mexico); Seidel, J.L. [ISTEEM, M.S.E. Montpellier (France)

    2001-07-01

    The radon behavior, chemical compounds, major and trace elements in water samples of four springs and three wells of urban and agricultural zones around the Jocotitlan volcano and El Oro region was determined, both of them located in the medium part of the Mexican neo-volcanic axis. The {sup 222} Rn was measured by the liquid scintillation method, the analysis of major components was realized with conventional chemical techniques, while the trace elements were quantified using an Icp-Ms. The average values of the radon concentrations obtained during one year were constant relatively, in an interval from 0.97 to 4.99 Bq/lt indicating a fast transport from the reload area toward the sampling points. the compounds, major and trace elements showed differences which indicate distinct origins of water from the site studies. (Author)

  4. Is the HIV epidemic stable among MSM in Mexico? HIV prevalence and risk behavior results from a nationally representative survey among men who have sex with men.

    Directory of Open Access Journals (Sweden)

    Sergio Bautista-Arredondo

    Full Text Available BACKGROUND: Recent evidence points to the apparent increase of HIV prevalence among men who have sex with men (MSM in different settings with concentrated epidemics, including the Latin American region. In 2011, Mexico implemented an ambitious HIV prevention program in all major cities, funded by the Global Fund to Fight Aids, Tuberculosis and Malaria. The program was intended to strengthen the prevention response for the most at risk populations: MSM and injecting drug users. This paper presents the HIV prevalence results of a nationally representative baseline survey in 24 Mexican cities throughout the 5 regions in the country and reports the socio-demographic and sexual risk behaviors that predict the probability of infection. METHODS: The survey was implemented in two phases. We first identified and characterized places where MSM gather in each city and then conducted in a second phase, a seroprevalence survey that included rapid HIV testing and a self-administered questionnaire. The prevalence of HIV was estimated by adjusting for positive predicted value. We applied a probit model to estimate the probability of having a positive result from the HIV test as a function of socio-demographic characteristics and self-reported sexual risk behaviors. RESULTS: We found an overall HIV prevalence among MSM gathering in meeting points of 16.9% [95% CI: 15.6-18.3], significantly higher than previously reported estimates. Our regression results suggest that the risk of infection increases with age, with the number of sexual partners, and among those who play a receptive sexual role, and the risk decreases with higher education. DISCUSSION: Our findings suggest a higher HIV prevalence among MSM than previously acknowledged and that a significant regional variability exist throughout the country. These two findings combined, signal an important dynamic in the epidemic that should be better understood and promptly addressed with strong prevention efforts

  5. Effect of pentachlorophenol and chemical oxygen demand mass concentrations in influent on operational behaviors of upflow anaerobic sludge blanket (UASB) reactor

    International Nuclear Information System (INIS)

    Shen Dongsheng; He Ruo; Liu Xinwen; Long Yan

    2006-01-01

    Upflow anaerobic sludge blanket (UASB) reactor that was seeded with anaerobic sludge acclimated to chlorophenols was used to investigate the feasibility of anaerobic biotreatment of synthetic wastewater containing pentachlorophenol (PCP) with additional sucrose as carbon source. Two sets of UASB reactors were operated at one time. But the seeded sludge for the two reactors was different and Reactor I was seeded with the sludge that was acclimated to PCP completely for half a year, and Reactor II was seeded with the mixed sludge that was acclimated for half a year to PCP, 4-CP, 3-CP or 2-CP, respectively. The degradation of PCP and the operation fee treating the wastewater are affected by the concentration of MEDS (microorganism easily degradable substrate). So the confirmation of the suitable ratio of [COD] and [PCP] was the key factor of treating the wastewater containing PCP economically and efficiently. During the experiment, the synthetic wastewater with 180.0 mg L -1 PCP and 1250-10000 mg L -1 COD could be treated steadily in the experimental Reactor I. The removal efficiency of PCP was more than 99.5% and the removal efficiency of COD was up to 90%. [PCP] (concentration of PCP) in effluent was less than 0.5 mg L -1 . [PCP] in influent could affect proper [COD] (concentration of COD) range in influent that was required for maintenance of steady running of the experimental reactor with a hydraulic retention time (HRT) from 20 to 22 h. [PCP] in influent would directly affect the necessary [COD] in influent when the UASB reactor ran normally and treated the wastewater containing PCP. When [PCP] was 100.4, 151.6 and 180.8 mg L -1 in influent, respectively, [COD] in influent had to be controlled about 1250-7500, 2500-5000 and 5000 mg L -1 to maintain the UASB reactor steady running normally and contemporarily ensure that [COD] and [PCP] in effluent were less than 300 and 0.5 mg L -1 , respectively. With the increase of [PCP] in influent, the range of variation of

  6. Effect of pentachlorophenol and chemical oxygen demand mass concentrations in influent on operational behaviors of upflow anaerobic sludge blanket (UASB) reactor.

    Science.gov (United States)

    Shen, Dong-Sheng; He, Ruo; Liu, Xin-Wen; Long, Yan

    2006-08-25

    Upflow anaerobic sludge blanket (UASB) reactor that was seeded with anaerobic sludge acclimated to chlorophenols was used to investigate the feasibility of anaerobic biotreatment of synthetic wastewater containing pentachlorophenol (PCP) with additional sucrose as carbon source. Two sets of UASB reactors were operated at one time. But the seeded sludge for the two reactors was different and Reactor I was seeded with the sludge that was acclimated to PCP completely for half a year, and Reactor II was seeded with the mixed sludge that was acclimated for half a year to PCP, 4-CP, 3-CP or 2-CP, respectively. The degradation of PCP and the operation fee treating the wastewater are affected by the concentration of MEDS (microorganism easily degradable substrate). So the confirmation of the suitable ratio of [COD] and [PCP] was the key factor of treating the wastewater containing PCP economically and efficiently. During the experiment, the synthetic wastewater with 180.0 mg L(-1) PCP and 1250-10000 mg L(-1) COD could be treated steadily in the experimental Reactor I. The removal efficiency of PCP was more than 99.5% and the removal efficiency of COD was up to 90%. [PCP] (concentration of PCP) in effluent was less than 0.5 mg L(-1). [PCP] in influent could affect proper [COD] (concentration of COD) range in influent that was required for maintenance of steady running of the experimental reactor with a hydraulic retention time (HRT) from 20 to 22 h. [PCP] in influent would directly affect the necessary [COD] in influent when the UASB reactor ran normally and treated the wastewater containing PCP. When [PCP] was 100.4, 151.6 and 180.8 mg L(-1) in influent, respectively, [COD] in influent had to be controlled about 1250-7500, 2500-5000 and 5000 mg L(-1) to maintain the UASB reactor steady running normally and contemporarily ensure that [COD] and [PCP] in effluent were less than 300 and 0.5 mg L(-1), respectively. With the increase of [PCP] in influent, the range of variation

  7. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

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

  8. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  9. stableGP

    Data.gov (United States)

    National Aeronautics and Space Administration — The code in the stableGP package implements Gaussian process calculations using efficient and numerically stable algorithms. Description of the algorithms is in the...

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  11. Reactor utilization

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-01-01

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

  12. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  13. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Araki, Takao; Saito, Yasushi.

    1987-01-01

    Purpose: To reduce the seismic wave responsivity of an exhaust duct shields thereby preventing the release of tritium in an evacuating device due to failures upon earthquakes. Constitution: The ends on the cutting side of upper outer exhaust duct shields of a thermonuclear reactor are connected with a plurality of support beams. In a case where seismic vibrations are exerted to such a thermonuclear reactor, since the ends on the cutting side are coupled with the support beams, vibrations of the upper outer exhaust duct shields are greatly restricted. Thus, since there is no more such a possibility, for example, that an exhaust duct connected to the upper portion of a reactor main body is greatly distorted due to the seismic response of the upper outside exhaust duct shields to result in the failure of the connection portion with a vacuum pump, the release of tritium due to failure of the evacuating device can be prevented. (Yoshino, Y.)

  14. Nuclear reactors design study and parameters calculation

    International Nuclear Information System (INIS)

    Morcos, H.N.

    2002-01-01

    the nuclear design a reactor core needs to determine a set of system parameters which will lead to safe, reliable and economical reactor operation at the rated power level over the desired core lifetime. the principal tools used in this task consist of a number of models of neutron behavior in the reactor that are implemented by a multiplicity of computer programs or codes used to simulate the nuclear behavior of the reactor core. the study of the interaction of the core power distributions with the time-dependent production or depletion of nuclei in the core is known as depletion or burn up analysis the main objective of the present thesis is to study the fuel depletion analysis under different reactor operating regimes and their influence on the build up of actinides and fission products (F P). therefore, one can estimate the optimum reactor-operating regime at which the accumulation of certain actinide isotope can reach maximum

  15. Population Games, Stable Games, and Passivity

    Directory of Open Access Journals (Sweden)

    Michael J. Fox

    2013-10-01

    Full Text Available The class of “stable games”, introduced by Hofbauer and Sandholm in 2009, has the attractive property of admitting global convergence to equilibria under many evolutionary dynamics. We show that stable games can be identified as a special case of the feedback-system-theoretic notion of a “passive” dynamical system. Motivated by this observation, we develop a notion of passivity for evolutionary dynamics that complements the definition of the class of stable games. Since interconnections of passive dynamical systems exhibit stable behavior, we can make conclusions about passive evolutionary dynamics coupled with stable games. We show how established evolutionary dynamics qualify as passive dynamical systems. Moreover, we exploit the flexibility of the definition of passive dynamical systems to analyze generalizations of stable games and evolutionary dynamics that include forecasting heuristics as well as certain games with memory.

  16. Model predictive control of a solar-thermal reactor

    Science.gov (United States)

    Saade Saade, Maria Elizabeth

    Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

  17. Simulation of a marine nuclear reactor

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki

    1995-01-01

    A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship's motions because of the ship's maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship's motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship's motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship's motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship's motions on the reactor behavior can be accurately simulated by NESSY

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  19. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  20. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  1. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  2. Neutronic reactor

    International Nuclear Information System (INIS)

    Carleton, J.T.

    1977-01-01

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment. 3 claims, 6 figures

  3. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells

    International Nuclear Information System (INIS)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R.

    2001-01-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm 2 , to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  4. Utilisation of thorium in reactors

    Science.gov (United States)

    Anantharaman, K.; Shivakumar, V.; Saha, D.

    2008-12-01

    India's nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th-Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle.

  5. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  7. Behavior to the fracture of an AISI 304 stainless steel sensitized in BWR reactor conditions (288 degrees Centigrade and 80 Kg/cm{sup 2}); Comportamiento a la fractura de un acero inoxidable AISI 304 sensibilizado en condiciones de reactor BWR (288 grados Centigrados y 80 Kg/cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Garcia R, R.; Aguilar T, A.; Gachuz M, M.; Arganis J, C.; Merino C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1999-07-01

    It is a knew fact that ductility of a lot of structural alloys can be deteriorated by the environment effect which are exposed, and that their consequent embrittlement can put in doubt the safety of their functioning; such is the case of austenitic stainless steels used in internal components of the BWR type reactors which not only is subjected to the effect combined of the aggressive environment which surround it (pressure, temperature, corrosion potential, conductivity medium, local state of efforts, etc.), but also to the action of present neutron radiation, manifesting microstructural changes which are reflected in the augmentation of its susceptibility to the intergranular cracking, phenomena generally known as IASCC ''Irradiation Assisted Stress Corrosion Cracking''. Once appeared the cracking in the material, the useful life of a component is limited by the rapidity to growth of these cracking, making necessary evaluations which can to predict its behavior, therefore the present work shows the preliminary results for determining the behavior to the fracture of an AISI 304 stainless steel sensitized, in a dynamic recirculation circuit which allows to simulate the operation conditions of a BWR reactor (288 Centigrade and 80 kg/cm{sup 2}). (Author)

  8. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  9. Trend of development of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, S. (Science and Technology Agency, Tokyo (Japan). Atomic Energy Bureau)

    1982-01-01

    The development of nuclear power is indispensable as the core of the substitute energy for petroleum. It is the urgent subject for world advanced countries to develop fast breeder reactors which can utilize uranium resources efficiently, to breed nuclear fuel resources, and to secure the stable supply of energy for long term in future. In Japan, the development of fast breeder reactors has been advanced independently and efficiently as a national project mainly by the Power Reactor and Nuclear Fuel Development Corp. The experimental reactor ''Joyo'' attained the criticality in April, 1977, and has been operated at the thermal output of 75,000 kW. As for the prototype reactor ''Monju'', the application for the permission to install it was submitted in December, 1980, and now, the safety examination is in progress. The present state of the development of fast breeder reactors in USA, Great Britain, France, West Germany, USSR and Japan is explained. In order to advance fast breeder reactors to the stage of full-scale practical use, a number of the reactors of 1 million kW class including the demonstration reactor will be constructed and operated to demonstrate and learn the technology of power generation plants in practical scale, to improve the performance, and to establish the economical efficiency. The schedule of development, the organization and the sharing of roles, the research and development in the demonstration stage are described.

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  12. Reactor container

    International Nuclear Information System (INIS)

    Otsuka, Hiroaki; Yoshida, Takashi.

    1979-01-01

    Purpose: To prevent rain water falling along the outer wall of the container during the construction work of an atomic power plant from making ingress into the inner part of a reactor container through a large size material carry-in port. Constitution: A weir for preventing the ingress of rain water is provided on the border between the foot floor of a large material carry-in port provided on the side surface at the bottom part of the reactor container and the floor surface of the building. This weir is of a semi-circular plate shape, and formed so that the lower semi-circular part of the carry-in port is tightly closed. (Kamimura, M.)

  13. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  14. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  15. The dismantling of fast reactors: sodium processing

    International Nuclear Information System (INIS)

    Rodriguez, G.; Berte, M.; Serpante, J.P.

    1999-01-01

    Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

  16. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  17. TRAP2: FORTRAN program for digital simulation of the transient behavior of the once-through steam generator and associated reactor coolant system

    International Nuclear Information System (INIS)

    Cudlin, J.J.; Daggett, P.W.

    1976-08-01

    The TRAP2 code calculates the transient response of the once-through steam generator and associated reactor coolant system to feedwater and main steam line pipe ruptures. The numerical integrations are made serially by a combination of forward differencing and implicit techniques. Fluid properties are obtained from equations that are surface fits to the 1967 ASME Steam Tables. Unusual features include multiple time-step analysis, restart capability, and plot capability

  18. Safety studies concerning nuclear power reactors

    International Nuclear Information System (INIS)

    Bailly, Jean; Pelce, Jacques

    1980-01-01

    The safety of nuclear installations poses different technical problems, whether concerning pressurized water reactors or fast reactors. But investigating methods are closely related and concern, on the one hand, the behavior of shields placed between fuel and outside and, on the other, analysis of accidents. The article is therefore in two parts based on the same plan. Concerning light water reactors, the programme of studies undertaken in France accounts for the research carried out in countries where collaboration agreements exist. Concerning fast reactors, France has the initiative of their studies owing to her technical advance, which explains the great importance of the programmes under way [fr

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  20. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  1. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    Science.gov (United States)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  2. Application of autoregressive methods and Lyapunov coefficients for instability studies of nuclear reactors

    International Nuclear Information System (INIS)

    Aruquipa Coloma, Wilmer

    2017-01-01

    Nuclear reactors are susceptible to instability, causing oscillations in reactor power in specific working regions characterized by determined values of power and coolant mass flow. During reactor startup, there is a greater probability that these regions of instability will be present; another reason may be due to transient processes in some reactor parameters. The analysis of the temporal evolution of the power reveals a stable or unstable process after the disturbance in a light water reactor of type BWR (Boiling Water Reactor). In this work, the instability problem was approached in two ways. The first form is based on the ARMA (Autoregressive Moving Average models) model. This model was used to calculate the Decay Ratio (DR) and natural frequency (NF) of the oscillations, parameters that indicate if the one power signal is stable or not. In this sense, the DRARMA code was developed. In the second form, the problems of instability were analyzed using the classical concepts of non-linear systems, such as Lyapunov exponents, phase space and attractors. The Lyapunov exponents quantify the exponential divergence of the trajectories initially close to the phase space and estimate the amount of chaos in a system; the phase space and the attractors describe the dynamic behavior of the system. The main aim of the instability phenomena studies in nuclear reactors is to try to identify points or regions of operation that can lead to power oscillations conditions. The two approaches were applied to two sets of signals. The first set comes from signals of instability events of the commercial Forsmark reactors 1 and 2 and were used to validate the DRARMA code. The second set was obtained from the simulation of transient events of the Peach Bottom reactor; for the simulation, the PARCS and RELAP5 codes were used for the neutronic/thermal hydraulic coupling calculation. For all analyzes made in this work, the Matlab software was used due to its ease of programming and

  3. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  4. System Requirements Document for the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  5. Stable isotopes labelled compounds

    International Nuclear Information System (INIS)

    1982-09-01

    The catalogue on stable isotopes labelled compounds offers deuterium, nitrogen-15, and multiply labelled compounds. It includes: (1) conditions of sale and delivery, (2) the application of stable isotopes, (3) technical information, (4) product specifications, and (5) the complete delivery programme

  6. Stable Boundary Layer Issues

    NARCIS (Netherlands)

    Steeneveld, G.J.

    2012-01-01

    Understanding and prediction of the stable atmospheric boundary layer is a challenging task. Many physical processes are relevant in the stable boundary layer, i.e. turbulence, radiation, land surface coupling, orographic turbulent and gravity wave drag, and land surface heterogeneity. The

  7. Evolutionary Stable Strategy

    Indian Academy of Sciences (India)

    IAS Admin

    After Maynard-Smith and Price [1] mathematically derived why a given behaviour or strategy was adopted by a certain proportion of the population at a given time, it was shown that a strategy which is currently stable in a population need not be stable in evolutionary time (across generations). Additionally it was sug-.

  8. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  9. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  10. Small reactors in the Canadian context: opportunities and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Walker, R.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This presentation discusses the opportunities and challenges for small reactors in Canada. It concludes by suggesting that the success of small reactors in Canada will depend on a number of factors including private sector investment, access to international markets, stable, equitable and adaptable regulatory regime, public trust and technology.

  11. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  12. Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor; Control difuso adaptable de la potencia neutronica del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.

    2014-07-01

    The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n{sub o} up to a predetermined final level n{sub f}. During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

  13. Damping of reactor internals

    International Nuclear Information System (INIS)

    Singleton, N.R.; Bohm, G.J.

    1977-01-01

    This paper presents and discusses the results of a study of internals damping using data obtained from wave analysis of PWR plant flow tests, and shaker tests. The damping values were obtained from vibration data taken during the pre-operational testing of several reactor plants and some in air shaker tests. Parameters available in the data include the presence of the core, the presence and position of the control rod drive line, reactor coolant temperature, and combination of reactor coolant pumps in operation. Modal damping values for the structures reported were obtained from the decay of autocorrellograms or from the modal response half-power bandwidths of frequency spectra. For the lower frequency core barrel-reactor vessel beam modes 2% to 5% damping values were found for minimum damping values. Significantly higher values are found in the data when, for example, intermittent contact occurs at the core barrel level supports. Core barrel and thermal shield shell modes having natural frequencies in the frequency range of interest for seismic response calculations exhibit damping values generally on the order of 1% to 2%. Higher frequency, very low amplitude, shell modes of these structures can have damping values of less than 1%. Damping values for guide tubes were found to have minimum values of 2% to 5% depending on their core location. The cross flow velocity and thus the floor turbulence excited amplitude is higher for guide tubes in core locations near the outlet nozzles. Information on the damping of upper support columns which are similarly excited is also given. Damping values reported are suitable for normal operation design conditions, i.e., for oscillatory behavior and relatively small amplitudes. The extrapolation of the data to obtain realistic values for large seismic events and for loss of coolant accidents is also discussed

  14. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

    A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  16. Normal modified stable processes

    DEFF Research Database (Denmark)

    Barndorff-Nielsen, Ole Eiler; Shephard, N.

    2002-01-01

    This paper discusses two classes of distributions, and stochastic processes derived from them: modified stable (MS) laws and normal modified stable (NMS) laws. This extends corresponding results for the generalised inverse Gaussian (GIG) and generalised hyperbolic (GH) or normal generalised inverse...... Gaussian (NGIG) laws. The wider framework thus established provides, in particular, for added flexibility in the modelling of the dynamics of financial time series, of importance especially as regards OU based stochastic volatility models for equities. In the special case of the tempered stable OU process...

  17. Applications of stable isotopes

    International Nuclear Information System (INIS)

    Letolle, R.; Mariotti, A.; Bariac, T.

    1991-06-01

    This report reviews the historical background and the properties of stable isotopes, the methods used for their measurement (mass spectrometry and others), the present technics for isotope enrichment and separation, and at last the various present and foreseeable application (in nuclear energy, physical and chemical research, materials industry and research; tracing in industrial, medical and agronomical tests; the use of natural isotope variations for environmental studies, agronomy, natural resources appraising: water, minerals, energy). Some new possibilities in the use of stable isotope are offered. A last chapter gives the present state and forecast development of stable isotope uses in France and Europe

  18. Comparison of neutronic behavior of UO2, (Th-233UO2 and (Th-235UO2 fuels in a typical heavy water reactor

    Directory of Open Access Journals (Sweden)

    Seyed Mohammad Mirvakili

    2015-04-01

    The obtained computational data showed both thorium-based fuels caused less negative temperature reactivity coefficients for the modeled research reactor in comparison with UO2 fuel loading. By contrast, 233U-containing thorium-based fuel and 235U-containing thorium-based fuel loadings in the thermal core did not drastically reduce the effective delayed neutron fractions and delayed neutron fractions compared to UO2 fuel. A provided higher conversion factor and lower transuranic production in the research core fed by the thorium-based fuels make the fuel favorable in achieving higher cycle length and less dangerous and costly nuclear disposals.

  19. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  20. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    International Nuclear Information System (INIS)

    Monteleone, S.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting

  1. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  2. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  3. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  4. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

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

  5. Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

    2016-09-26

    In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

  6. Analysing Stable Time Series

    National Research Council Canada - National Science Library

    Adler, Robert

    1997-01-01

    We describe how to take a stable, ARMA, time series through the various stages of model identification, parameter estimation, and diagnostic checking, and accompany the discussion with a goodly number...

  7. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  8. Reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-09-07

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.).

  9. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  10. Particulate behavior in a controlled-profile pulverized coal-fired reactor: A study of coupled turbulent particle dispersion and thermal radiation transport. Final technical progress report

    Energy Technology Data Exchange (ETDEWEB)

    Queiroz, M.; Webb, B.W.

    1996-06-01

    To aid in the evaluation and development of advanced coal-combustion models, comprehensive experimental data sets are needed containing information on both the condensed and gas phases. To address this need a series of test were initiated on a 300 kW laboratory-scale, coal-fired reactor at a single test condition using several types of instrumentation. Data collected on the reactor during the course of the test includes: gas, particle, and wall temperature profiles; radiant, total, and convective heat fluxes to the walls; particle size and velocity profiles; transmission measurements; and gas species concentrations. Solid sampling was also performed to determine carbon and total burnout. Along with the extensive experimental measurements, the particle dispersion and radiation submodels in the ACERC comprehensive 2D code were studied in detail and compared to past experimental measurements taken in the CPR. In addition to the presentation and discussion of the experimental data set, a detailed description of the measurement techniques used in collecting the data, including a discussion of the error associated with each type of measurement, is given.

  11. Irradiation behavior of developed radiation resistance optical-fibers and observed optical radiation from their SiO2 cores under reactor irradiation

    International Nuclear Information System (INIS)

    Shikama, Tatsuo; Narui, Minoru; Kayano, Hideo; Kakuta, Tsunemi; Sagawa, Tsutomu; Sanada, Kazuo; Shamoto, Naoki; Uramoto, Toshimasa.

    1994-01-01

    Two kinds of optical fibers were irradiated in a fission reactor, JMTR(Japan Materials Testing Reactor), up to a 1.55x10 19 n/cm 2 fast neutron fluence and a 3.3x10 9 Gy ionizing dose at 370K. Optical transmission spectra were measured in the wavelength range of 450-1750nm, in-situ. Growth of strong optical absorption bands were observed in the range of wavelength shorter than 750nm. In the meantime, the fibers showed good radiation-resistance in the range of wavelength longer than 750nm. Optical radiations were observed from SiO 2 optical fibers under irradiation. A major part of the observed optical radiations is thought to be composed of broad optical radiation in the whole wavelength range studied in the present experiment. This broad optical radiation will be generated by the process of so-called Cerenkov radiation. Also, a sharp optical radiation peak was found at 1270nm on a F-doped fiber. This peak is thought to relate with doped Fluorine ions and ionizing gamma-ray irradiation. (author)

  12. Meteorological aspects of the reactor safety study requiring further study

    International Nuclear Information System (INIS)

    Slinn, W.G.N.

    1981-01-01

    Simple and approximate methods are used in a search for meteorological features that dominate estimates of reactor-accident consequences, and that require more accurate descriptions if consequence estimates are to be more realistic. By considering variations in the source term, it is seen that accidents involving containment-vessel failure dominate both the mean and variance of the distribution of consequences, although this conclusion is subject to uncertainties about plume rise. Research is recommended on the behavior of horizontal, sonic jets, with heat transfer to the ground, and especially during stable atmospheric conditions. Diffusion with fumigation and lofting require further study; use of K-theory and National Weather Service data should be vigorously pursued. Conditional upon an accident occurring, precipitation scavenging appears to dominate the variance of the consequences

  13. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    International Nuclear Information System (INIS)

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

    2004-01-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

  14. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core; Comportamento termofluidodinamico do gas refrigerante helio em um canal topico de reator VHTGR de nucleo prismatico

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Allan Cavalcante

    2016-08-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4{sup th} generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  16. NEUTRONIC REACTOR

    Science.gov (United States)

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  18. A study on the in-reactor behavior of centrifugally atomized U3Si dispersion fuel irradiated up to high burn-up under normal power condition

    International Nuclear Information System (INIS)

    Kim, K. H.; Lee, Y. S.; Lee, C. S.; Jeong, Y. H.; Park, D. G.; Kim, C. G.

    2001-01-01

    In order to localize the nuclear fuels for HANARO, centrifugally atomized and mechanically comminuted U 3 Si mini-element fuels were fabricated and irradiated up to high burn-up under normal power condition in HANARO reactor. The average interaction thickness in atomized U 3 Si dispersion fuel meat was observed as about 8 μm. It was almost same thickness and more uniform comparing to that of comminuted U 3 Si dispersion fuel meats. The interaction thickness of U 3 Si dispersion fuel meats were generally assumed acceptable in the fuel performance point of view. The U 3 Si fuel particles were examined containing a relatively fine and uniform fission gas bubble distribution without any indications of unstable and rapid breakway swelling, irrespective of powdering methods

  19. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  20. Improvements in streaking nuclear reactors

    International Nuclear Information System (INIS)

    Pedrick, A.P.

    1976-01-01

    In this type of reactor atomic nuclei are stripped of their electron shells by heating to form a very high temperature plasma which is passed at high speed through a chamber in which they are forced into contact with a 'wall' formed by a unidirectional stream of photons from continuous laser beams. In this way it should be possible to brush off from the surface of the nuclei protons and neutrons, with release of their binding energy. The energy thus produced can be subjected to much more gentle control than with a fission or fusion reactor. Furthermore, if this concept can be successfully applied to elements of high atomic number which are normally regarded as stable and unfissionable, a vast new source of nuclear energy release will have been made available. It also seems possible that an atomic nucleus might be spun sufficiently in such a reactor to disintegrate it completely into nucleons by simple centrifugal action, with great release of binding energy. The reactor described has a central body with radial ducts through which the nuclei are passed, and a number of lasers are provided whose beams are arranged so that the nuclei are discharged at the cross-over point of two or more laser beams which form a corner at the junction of two or more photon walls. (U.K.)

  1. Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

    International Nuclear Information System (INIS)

    Popa-Simil, Liviu

    2008-01-01

    Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

  2. Calcium stable isotope geochemistry

    Energy Technology Data Exchange (ETDEWEB)

    Gausonne, Nikolaus [Muenster Univ. (Germany). Inst. fuer Mineralogie; Schmitt, Anne-Desiree [Strasbourg Univ. (France). LHyGeS/EOST; Heuser, Alexander [Bonn Univ. (Germany). Steinmann-Inst. fuer Geologie, Mineralogie und Palaeontologie; Wombacher, Frank [Koeln Univ. (Germany). Inst. fuer Geologie und Mineralogie; Dietzel, Martin [Technische Univ. Graz (Austria). Inst. fuer Angewandte Geowissenschaften; Tipper, Edward [Cambridge Univ. (United Kingdom). Dept. of Earth Sciences; Schiller, Martin [Copenhagen Univ. (Denmark). Natural History Museum of Denmark

    2016-08-01

    This book provides an overview of the fundamentals and reference values for Ca stable isotope research, as well as current analytical methodologies including detailed instructions for sample preparation and isotope analysis. As such, it introduces readers to the different fields of application, including low-temperature mineral precipitation and biomineralisation, Earth surface processes and global cycling, high-temperature processes and cosmochemistry, and lastly human studies and biomedical applications. The current state of the art in these major areas is discussed, and open questions and possible future directions are identified. In terms of its depth and coverage, the current work extends and complements the previous reviews of Ca stable isotope geochemistry, addressing the needs of graduate students and advanced researchers who want to familiarize themselves with Ca stable isotope research.

  3. Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.

    2014-01-01

    The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n o up to a predetermined final level n f . During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

  4. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

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

  5. Survey of research reactors

    International Nuclear Information System (INIS)

    Boek, H.; Villa, M.

    2004-06-01

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

  6. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  7. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Hagrman, D.L.; Reymann, G.A.

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

  8. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.L.; Reymann, G.A. (comps.)

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  9. Evolutionary Stable Strategy

    Indian Academy of Sciences (India)

    2016-08-26

    Aug 26, 2016 ... Home; Journals; Resonance – Journal of Science Education; Volume 21; Issue 9. Evolutionary Stable Strategy: Application of Nash Equilibrium in Biology. General ... Using some examples of classical games, we show how evolutionary game theory can help understand behavioural decisions of animals.

  10. Manifolds admitting stable forms

    Czech Academy of Sciences Publication Activity Database

    Le, Hong-Van; Panák, Martin; Vanžura, Jiří

    2008-01-01

    Roč. 49, č. 1 (2008), s. 101-11 ISSN 0010-2628 R&D Projects: GA ČR(CZ) GP201/05/P088 Institutional research plan: CEZ:AV0Z10190503 Keywords : stable forms * automorphism groups Subject RIV: BA - General Mathematics

  11. Stable isotope studies

    International Nuclear Information System (INIS)

    Ishida, T.

    1992-01-01

    The research has been in four general areas: (1) correlation of isotope effects with molecular forces and molecular structures, (2) correlation of zero-point energy and its isotope effects with molecular structure and molecular forces, (3) vapor pressure isotope effects, and (4) fractionation of stable isotopes. 73 refs, 38 figs, 29 tabs

  12. Interactive Stable Ray Tracing

    DEFF Research Database (Denmark)

    Dal Corso, Alessandro; Salvi, Marco; Kolb, Craig

    2017-01-01

    Interactive ray tracing applications running on commodity hardware can suffer from objectionable temporal artifacts due to a low sample count. We introduce stable ray tracing, a technique that improves temporal stability without the over-blurring and ghosting artifacts typical of temporal post-pr...

  13. The stable subgroup graph

    Directory of Open Access Journals (Sweden)

    Behnaz Tolue

    2018-07-01

    Full Text Available In this paper we introduce stable subgroup graph associated to the group $G$. It is a graph with vertex set all subgroups of $G$ and two distinct subgroups $H_1$ and $H_2$ are adjacent if $St_{G}(H_1\\cap H_2\

  14. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  15. Determining Reactor Neutrino Flux

    OpenAIRE

    Cao, Jun

    2011-01-01

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understa...

  16. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  18. Reactor container

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo; Yanai, Ryoichi.

    1976-01-01

    Object: To provide a reactor container which is free from water shock action or condensing vibrations and cannot be readily broken by a missile from a pump impeller, pipe whipping, steam jet reaction, etc., and which also quickly condenses issuing steam and possesses a large vibration-proof strength. Structure: A high pressure containment vessel accommodating a pressure container includes a plurality of pressurized water tanks arranged along its inner periphery, and a pneumatic valve is provided in a lower portion of each of these pressurized water tanks. If an accident occurs, vapor is caused to issue from the pressure container into the vessel. When a certain value is reached, the pneumatic valves are opened, whereby the gas within the pressurized water tanks causes pressurized water to flow through the pipe and be ejected from spray nozzles to cause condensation of water within the vessel. Further, water of a pool within the container is circulated to allow heat release to the outside. (Horiuchi, T.)

  19. Fast pyrolysis of durian (Durio zibethinus L) shell in a drop-type fixed bed reactor: Pyrolysis behavior and product analyses.

    Science.gov (United States)

    Tan, Y L; Abdullah, A Z; Hameed, B H

    2017-11-01

    Durian shell (DS) was pyrolyzed in a drop-type fixed-bed reactor to study the physicochemical properties of the products. The experiment was carried out with different particle sizes (up to 5mm) and reaction temperatures (250-650°C). The highest bio-oil yield was obtained at 650°C (57.45wt%) with DS size of 1-2mm. The elemental composition and higher heating value of the feedstock, bio-oil (650°C), and bio-char (650°C) were determined and compared. The compositions of product gases were determined via gas chromatography with thermal conductivity detector. The chemical composition of bio-oil was analyzed by gas chromatography-mass spectrometry. The bio-oil produced at lower temperature yields more alcohols, whereas the bio-oil produced at higher temperature contains more aromatics and carbonyls. Bio-oil has potential to be used as liquid fuel or fine chemical precursor after further upgrading. The results further showed the potential of bio-char as a solid fuel. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  1. Reactor shutdown device

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Yuji; Sakamoto, Sadao.

    1992-01-27

    In control rod drives, the attracting portion between an electromagnet and a magnetic material is formed into a shape engagable to each other, and a protrusion is disposed at the outer circumferential surface. In addition, a recess is disposed at the lower portion of the electromagnet, and a protrusion to be inserted into the recess is formed at the upper portion of the magnetic member. A universal coupling is disposed at the upper portion of a control rod, and when horizontal vibrations are caused by earthquakes or by the behavior of coolants, the horizontal vibrations are moderated by the universal coupling. In this structure, since the attracting portion between the electromagnet and the magnetic member is formed into a shape like that of a receptable structure and the attracting portion is always brought into contact with each other, shearing loads are extremely reduced. Accordingly, a stress tending to separate the attracting portion is not exerted even upon occurrence of earthquakes. That is, it is possible to prevent erroneous operation of the reactor scram performance attributable to causes other than normal scram operation of dropping control rods by positively eliminating the attracting force of the electromagnet to conduct reactor scram. (N.H.).

  2. Preliminary design of a Binary Breeder Reactor

    International Nuclear Information System (INIS)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C.

    2014-10-01

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  3. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  4. Improved reactor regulating system logical architecture using genetic algorithm

    Directory of Open Access Journals (Sweden)

    Hyo-Sub Shim

    2017-12-01

    Full Text Available An improved Reactor Regulating System (RRS logic architecture, which is combined with genetic algorithm (GA, is implemented in this work. It is devised to provide an optimal solution to the current RRS. The current system works desirably and has contributed to safe and stable nuclear power plant operation. However, during the ascent and descent section of the reactor power, the RRS output reveals a relatively high steady-state error, and the output also carries a considerable level of overshoot. In an attempt to consolidate conservatism and minimize the error, this work proposes to apply GA to RRS and suggests reconfiguring the system. Prior to the use of GA, reverse engineering is implemented to build a Simulink-based RRS model. Reengineering is followed to produce a newly configured RRS to generate an output that has a reduced steady-state error and diminished overshoot level. A full-scope APR1400 simulator is used to examine the dynamic behaviors of RRS and to build the RRS Simulink model.

  5. Development of a general coupling interface for the fuel performance code TRANSURANUS – Tested with the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.; Macián-Juan, R.

    2015-01-01

    Highlights: • A general coupling interface was developed for couplings of the TRANSURANUS code. • With this new tool simplified fuel behavior models in codes can be replaced. • Applicable e.g. for several reactor types and from normal operation up to DBA. • The general coupling interface was applied to the reactor dynamics code DYN3D. • The new coupled code system DYN3D–TRANSURANUS was successfully tested for RIA. - Abstract: A general interface is presented for coupling the TRANSURANUS fuel performance code with thermal hydraulics system, sub-channel thermal hydraulics, computational fluid dynamics (CFD) or reactor dynamics codes. As first application the reactor dynamics code DYN3D was coupled at assembly level in order to describe the fuel behavior in more detail. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach transfers parameters like fuel temperature and cladding temperature back to DYN3D. Results of the coupled code system are presented for the reactivity transient scenario, initiated by control rod ejection. More precisely, the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy. These differences can be explained thanks to the greater detail in fuel behavior modeling. The numerical performance for DYN3D–TRANSURANUS was proved to be fast and stable. The coupled code system can therefore improve the assessment of safety criteria, at a reasonable computational cost

  6. OECD: Halden reactor project

    International Nuclear Information System (INIS)

    1979-01-01

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

  7. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

    2016-01-01

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  8. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  9. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  10. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    Energy Technology Data Exchange (ETDEWEB)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  11. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  12. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  13. Stable isotope analysis

    International Nuclear Information System (INIS)

    Tibari, Elghali; Taous, Fouad; Marah, Hamid

    2014-01-01

    This report presents results related to stable isotopes analysis carried out at the CNESTEN DASTE in Rabat (Morocco), on behalf of Senegal. These analyzes cover 127 samples. These results demonstrate that Oxygen-18 and Deuterium in water analysis were performed by infrared Laser spectroscopy using a LGR / DLT-100 with Autosampler. Also, the results are expressed in δ values (‰) relative to V-SMOW to ± 0.3 ‰ for oxygen-18 and ± 1 ‰ for deuterium.

  14. Forensic Stable Isotope Biogeochemistry

    Science.gov (United States)

    Cerling, Thure E.; Barnette, Janet E.; Bowen, Gabriel J.; Chesson, Lesley A.; Ehleringer, James R.; Remien, Christopher H.; Shea, Patrick; Tipple, Brett J.; West, Jason B.

    2016-06-01

    Stable isotopes are being used for forensic science studies, with applications to both natural and manufactured products. In this review we discuss how scientific evidence can be used in the legal context and where the scientific progress of hypothesis revisions can be in tension with the legal expectations of widely used methods for measurements. Although this review is written in the context of US law, many of the considerations of scientific reproducibility and acceptance of relevant scientific data span other legal systems that might apply different legal principles and therefore reach different conclusions. Stable isotopes are used in legal situations for comparing samples for authenticity or evidentiary considerations, in understanding trade patterns of illegal materials, and in understanding the origins of unknown decedents. Isotope evidence is particularly useful when considered in the broad framework of physiochemical processes and in recognizing regional to global patterns found in many materials, including foods and food products, drugs, and humans. Stable isotopes considered in the larger spatial context add an important dimension to forensic science.

  15. Stable SU(5) monopoles with higher magnetic charge

    International Nuclear Information System (INIS)

    Miyamoto, S.; Sato, H.; Tomohiro, S.

    1985-01-01

    Taking into account the electroweak breaking effects, some multiply charged monopoles were shown to be stable by Gardner and Harvey. We give the explicit Ansa$uml: tze for finite-energy, nonsingular solutions of these stable higher-strength monopoles with eg = 1,(3/2),3. We also give the general stability conditions and the detailed behavior of the interaction potentials between two monopoles which produce the stable higher-strength monopoles

  16. Local Search Approaches in Stable Matching Problems

    Directory of Open Access Journals (Sweden)

    Toby Walsh

    2013-10-01

    Full Text Available The stable marriage (SM problem has a wide variety of practical applications, ranging from matching resident doctors to hospitals, to matching students to schools or, more generally, to any two-sided market. In the classical formulation, n men and n women express their preferences (via a strict total order over the members of the other sex. Solving an SM problem means finding a stable marriage where stability is an envy-free notion: no man and woman who are not married to each other would both prefer each other to their partners or to being single. We consider both the classical stable marriage problem and one of its useful variations (denoted SMTI (Stable Marriage with Ties and Incomplete lists where the men and women express their preferences in the form of an incomplete preference list with ties over a subset of the members of the other sex. Matchings are permitted only with people who appear in these preference lists, and we try to find a stable matching that marries as many people as possible. Whilst the SM problem is polynomial to solve, the SMTI problem is NP-hard. We propose to tackle both problems via a local search approach, which exploits properties of the problems to reduce the size of the neighborhood and to make local moves efficiently. We empirically evaluate our algorithm for SM problems by measuring its runtime behavior and its ability to sample the lattice of all possible stable marriages. We evaluate our algorithm for SMTI problems in terms of both its runtime behavior and its ability to find a maximum cardinality stable marriage. Experimental results suggest that for SM problems, the number of steps of our algorithm grows only as O(n log(n, and that it samples very well the set of all stable marriages. It is thus a fair and efficient approach to generate stable marriages. Furthermore, our approach for SMTI problems is able to solve large problems, quickly returning stable matchings of large and often optimal size, despite the

  17. Autonomous Control of Space Nuclear Reactors

    Science.gov (United States)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  18. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    fuels as well as the applied methodologies. The IRSN proceeds in a relevant and independent assessment of the submitted safety reports. To achieve this goal and maintain over time an independent and relevant assessment capability, the IRSN relies on the excellence of its experts and on state of art techniques and knowledge. The IRSN contributes by its work in key area to cutting edge research and development in order to drive nuclear industry towards making the best use of scientific and technological progress for improving safety, environmental protection and health. To maintain at all times the state of the art knowledge and the operational expertise necessary to deal efficiently with major nuclear accident consequences, the IRSN carries out, on the one hand, its own research and development programs to gain accurate knowledge on still unknown phenomena for safety analysis. On the other hand, the IRSN works out its own scientific calculation methodologies involving industrial calculation chain. Concerning more particularly the 'two-phase flows' thematic, The ISRN must correctly simulate the primary fluid behavior in the reactor in normal operation as well as in accidental situations, to estimate if, in such situations, the core reactor state is fully safe and any safety risk is undergone The research and development programs launched at the ISRN on two-phase flows gather work on advanced thermohydraulic configurations encounter in various reactor states (normal operation or accidental situations), in particular: (i)The estimation of the margin to the critical heat flux in normal operation (DNBR), (ii) The pressurized thermal shock, which is due to mechanical important constraints in the reactor vessel resulting from the injection of a cold fluid in case of emergency cooling (PTS), (iii) The reactivity insertion accident (RIA), (iv) The loss of coolant accident (LOCA), (vi) The accidents in spent-fuel pools and (vii) The severe accident, which could lead to core

  19. Multiphysics Model Development and the Core Analysis for In Situ Breeding and Burning Reactor

    Directory of Open Access Journals (Sweden)

    Shengyi Si

    2013-01-01

    Full Text Available The in situ breeding and burning reactor (ISBBR, which makes use of the outstanding breeding capability of metallic pellet and the excellent irradiation-resistant performance of SiCf/SiC ceramic composites cladding, can approach the design purpose of ultralong cycle and ultrahigh burnup and maintain stable radial power distribution during the cycle life without refueling and shuffling. Since the characteristics of the fuel pellet and cladding are different from the traditional fuel rod of ceramic pellet and metallic cladding, the multiphysics behaviors in ISBBR are also quite different. A computer code, named TANG, to model the specific multiphysics behaviors in ISBBR has been developed. The primary calculation results provided by TANG demonstrate that ISBBR has an excellent comprehensive performance of GEN-IV and a great development potential.

  20. Introduction to Safety Analysis Approach for Research Reactors

    International Nuclear Information System (INIS)

    Park, Suki

    2016-01-01

    The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

  1. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

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

  2. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

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

  3. The Oklo natural reactor, Gabon

    International Nuclear Information System (INIS)

    Brookins, D.G.

    1989-01-01

    This paper reports how the Oklo natural reactor, Gabon, remains the best natural analog for assessing the behavior of fission products, actinides, and actinide daughters in rocks. The rocks at Oklo are porous and permeable, as well as being fractured and containing abundant water, yet many of the fission products and actinides have remained in place or close to their formation sites. The actinides Th, U, Np, Pu, and Am are similar in their crystal chemical characteristics, and all were retained in the host pitchblende. Elements incompatible in the pitchblende structure were lost by diffusion into the rocks surrounding the high grade reactor ore. Alkali and Alkaline earth elements Rb, Sr, Cs, and Ba were fixed in and very close to the reactor ores in clay minerals and in some secondary carbonates and sulfates. Local oxidizing conditions in the reactor zones caused some loss of Tc, Mo, Cd, and Ru, but the Tc, Mo, and Ru were fixed in sulfides formed close by under chemically reducing conditions. Local migration for Ag and Sn has been documented for some samples

  4. 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

  5. Research reactor instrumentation

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2001-02-01

    This is a textbook on research reactor instrumentation for training purposes, it gives a survey on research reactor instrumentation requirements and eight exercises covering the major aspects of this topic are presented. (author)

  6. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  7. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  8. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

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

  9. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  10. Reactor shutdown method

    International Nuclear Information System (INIS)

    Nishino, Yoshitaka; Sawa, Toshio; Matsumoto, Takayuki; Osumi, Katsumi; Usui, Naoshi.

    1991-01-01

    A device for injecting a hydrogen gas, a chelating agent or a reducing agent is disposed in a reactor water recycling system. Upon reactor shutdown, the hydrogen gas, the chelating agent or the reducing agent is injected to primary coolants. With such a procedure, radioactive ions formed by the dissolution of oxide layers at the surface of pipelines and equipments in a reactor water recycling system and a reactor water cleanup system are removed from the primary coolants by a reactor water cleanup device. Accordingly, since the dose rate at the surface of the pipelines can be reduced, the operator's radiation dose can be reduced upon periodical inspection for a power plant. Further, the inner pressure of the reactor is kept higher than the saturated steam pressure at the reactor water temperature to suppress boiling of the reactor water. This can suppress the peeling of cruds deposited to the surface of the fuel cladding tube. (I.N.)

  11. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

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

  12. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

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

  13. Reactor physics in support of the naval nuclear propulsion programme

    International Nuclear Information System (INIS)

    Lisley, P.G.; Beeley, P.A.

    1994-01-01

    Reactor physics is a core component of all courses but in particular two postgraduate courses taught at the department in support of the naval nuclear propulsion programme. All of the courses include the following elements: lectures and problem solving exercises, laboratory work, experiments on the Jason zero power Argonaut reactor, demonstration of PWR behavior on a digital computer simulator and project work. This paper will highlight the emphasis on reactor physics in all elements of the education and training programme. (authors). 9 refs

  14. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  15. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  16. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  17. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1974-01-01

    A major part of the current research programme is devoted to irradiation experiments with a wide variety of heavily instrumented test fuel assemblies, in order to study the thermal and mechanical behavior of fuel rods through in-core measurements, in particular various forms of deformation of cladding and fuel as related to operational conditions and fuel rod design parameters. From these measurements mathematical models are being developed to explain quantitatively the deformation behavior, as well as the thermal properties of the fuel. During 1974, fifty-six instrumented fuel assemblies were irradiated in these experiments. Another major part of the Halden programme is aimed at the development and demonstration of advanced computer-based methods for plant and reactor core control, for safety and protection, and for overall supervision of nuclear power stations. Both the control methods themselves and the associated measurement and control apparatus are being elaborated, and during the year particular progress was made with the ''OPCOM'' process operator communication system

  18. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

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

  19. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

    Bouquin, B.; Rio, I.; Safieh, J.

    1997-01-01

    On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

  20. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  1. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  2. The Jules Horowitz reactor

    International Nuclear Information System (INIS)

    2003-01-01

    The Jules Horowitz reactor is the future european reactor for irradiation. It will be used for materials and new fuels irradiation. Experiments for the safety and the validation of neutronics calculation will be also realized. This paper presents the design and the performance of the reactor and the schedule of the remaining design studies. (A.L.B.)

  3. Thermally Stable, Latent Olefin Metathesis Catalysts

    OpenAIRE

    Thomas, Renee M.; Fedorov, Alexey; Keitz, Benjamin K.; Grubbs, Robert H.

    2011-01-01

    Highly thermally stable N-aryl,N-alkyl N-heterocyclic carbene (NHC) ruthenium catalysts were designed and synthesized for latent olefin metathesis. These catalysts showed excellent latent behavior toward metathesis reactions, whereby the complexes were inactive at ambient temperature and initiated at elevated temperatures, a challenging property to achieve with second generation catalysts. A sterically hindered N-tert-butyl substituent on the NHC ligand of the ruthenium complex was found to i...

  4. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  5. The development of fast reactors in France

    International Nuclear Information System (INIS)

    Vautrey, L.

    1982-01-01

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

  6. Marginally Stable Nuclear Burning

    Science.gov (United States)

    Strohmayer, Tod E.; Altamirano, D.

    2012-01-01

    Thermonuclear X-ray bursts result from unstable nuclear burning of the material accreted on neutron stars in some low mass X-ray binaries (LMXBs). Theory predicts that close to the boundary of stability oscillatory burning can occur. This marginally stable regime has so far been identified in only a small number of sources. We present Rossi X-ray Timing Explorer (RXTE) observations of the bursting, high- inclination LMXB 4U 1323-619 that reveal for the first time in this source the signature of marginally stable burning. The source was observed during two successive RXTE orbits for approximately 5 ksec beginning at 10:14:01 UTC on March 28, 2011. Significant mHz quasi- periodic oscillations (QPO) at a frequency of 8.1 mHz are detected for approximately 1600 s from the beginning of the observation until the occurrence of a thermonuclear X-ray burst at 10:42:22 UTC. The mHz oscillations are not detected following the X-ray burst. The average fractional rms amplitude of the mHz QPOs is 6.4% (3 - 20 keV), and the amplitude increases to about 8% below 10 keV.This phenomenology is strikingly similar to that seen in the LMXB 4U 1636-53. Indeed, the frequency of the mHz QPOs in 4U 1323-619 prior to the X-ray burst is very similar to the transition frequency between mHz QPO and bursts found in 4U 1636-53 by Altamirano et al. (2008). These results strongly suggest that the observed QPOs in 4U 1323-619 are, like those in 4U 1636-53, due to marginally stable nuclear burning. We also explore the dependence of the energy spectrum on the oscillation phase, and we place the present observations within the context of the spectral evolution of the accretion-powered flux from the source.

  7. Thorium fueled reactor

    Science.gov (United States)

    Sipaun, S.

    2017-01-01

    Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

  8. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  9. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  10. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1981-12-01

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

  11. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  12. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  13. Spinning fluids reactor

    Science.gov (United States)

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

    2012-11-20

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

  14. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  15. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  16. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  17. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  18. Advanced thermally stable jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.

    1999-01-31

    The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

  19. Analysis of LWR-MOX fuel steady behavior and code validation

    International Nuclear Information System (INIS)

    He Xiaojun; Chen Peng; Huang Yucai

    2010-01-01

    In this study, the fuel performance code METEOR v1.9 developed by CEA (France) was used to analyze the fuel behavior of MOX fuel of light water reactor from OECD-Halden project IFA597.4-7 test under steady state irradiation. The code predictions were compared with the experimental data. The research showed that: 1) The MOX fuel of light water reactor had stable performance under steady state irradiation condition. The performance of MOX fuel had no significant difference with UO 2 fuel, except that its fission gas release (FGR) was a little bit higher than UO 2 fuel; 2) METEOR1.9 code could predict behavior of MOX very well. The predictions of temperature distribution, rod inner pressure and fission gas release fraction had a good agreement with experimental data. (authors)

  20. APPLICATION OF MODEL PREDICTIVE CONTROL TO BATCH POLYMERIZATION REACTOR

    Directory of Open Access Journals (Sweden)

    N.M. Ghasem

    2006-06-01

    Full Text Available The absence of a stable operational state in polymerization reactors that operates in batches is factor that determine the need of a special control system. In this study, advanced control methodology is implemented for controlling the operation of a batch polymerization reactor for polystyrene production utilizingmodel predictive control. By utilizing a model of the polymerization process, the necessary operational conditions were determined for producing the polymer within the desired characteristics. The maincontrol objective is to bring the reactor temperature to its target temperature as rapidly as possible with minimal temperature overshoot. Control performance for the proposed method is encouraging. It has been observed that temperature overshoot can be minimized by the proposed method with the use of both reactor and jacket energy balance for reactor temperature control.

  1. Analysis and evaluation of the Dual Fluid Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiang

    2017-06-27

    The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

  2. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    The invention deals with disengaging the coupling of a reactor coolant pump of a nuclear reactor feeding pressurized coolant. The disengaging coupling has two parts joined by bolts, at least one of them containing a driving agent within a bore. This is provided with a speed-depending ignition device in such manner that, if the critical speed is reached, the driving charge is ignited and the coupling is disengaged by destroying the bolts. (UWI) [de

  3. Neutron fluxes in test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  4. A Course in Chemical Reactor Design.

    Science.gov (United States)

    Takoudis, Christos G.

    1983-01-01

    Presents course outline, topics covered, and final project (doubling as a take home final exam) for a one-semester, interdisciplinary course on the design and behavior of chemical reactors. Interplay of chemical and physical rate processes is stressed in the course. (JM)

  5. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  6. Wide band characterization of wind turbine reactors

    DEFF Research Database (Denmark)

    Holdyk, Andrzej; Arana Aristi, Ivan; Holbøll, Joachim

    2013-01-01

    This paper presents the results of field measure-ments of the impedance of two commercial available wind turbine reactors, performed by a sweep frequency response analyzer, sFRA. The measurements were taken in the frequency range from 20Hz to 20MHz. Comparable frequency behavior was found in all ...

  7. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  8. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  9. System Requirements Document for the Molten Salt Reactor Experiment 233U conversion system

    International Nuclear Information System (INIS)

    Aigner, R.D.

    2000-01-01

    The purpose of the conversion process is to convert the 233 U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019

  10. Characterization of methane oxidation in a simulated landfill cover system by comparing molecular and stable isotope mass balances.

    Science.gov (United States)

    Schulte, Marcel; Jochmann, Maik A; Gehrke, Tobias; Thom, Andrea; Ricken, Tim; Denecke, Martin; Schmidt, Torsten C

    2017-11-01

    Biological methane oxidation may be regarded as a method of aftercare treatment for landfills to reduce climate relevant methane emissions. It is of social and economic interest to estimate the behavior of bacterial methane oxidation in aged landfill covers due to an adequate long-term treatment of the gas emissions. Different approaches assessing methane oxidation in laboratory column studies have been investigated by other authors recently. However, this work represents the first study in which three independent approaches, ((i) mass balance, (ii) stable isotope analysis, and (iii) stoichiometric balance of product (CO 2 ) and reactant (CH 4 ) by CO 2 /CH 4 -ratio) have been compared for the estimation of the biodegradation by a robust statistical validation on a rectangular, wide soil column. Additionally, an evaluation by thermal imaging as a potential technique for the localization of the active zone of bacterial methane oxidation has been addressed in connection with stable isotope analysis and CO 2 /CH 4 -ratios. Although landfills can be considered as open systems the results for stable isotope analysis based on a closed system correlated better with the mass balance than calculations based on an open system. CO 2 /CH 4 -ratios were also in good agreement with mass balance. In general, highest values for biodegradation were determined from mass balance, followed by CO 2 /CH 4 -ratio, and stable isotope analysis. The investigated topsoil proved to be very suitable as a potential cover layer by removing up to 99% of methane for CH 4 loads of 35-65gm -2 d -1 that are typical in the aftercare phase of landfills. Finally, data from stable isotope analysis and the CO 2 /CH 4 -ratios were used to trace microbial activity within the reactor system. It was shown that methane consumption and temperature increase, as a cause of high microbial activity, correlated very well. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Development and verification of the FP behavior evaluation computer code: FPRETAIN

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1989-06-01

    This paper describes results of development and verification study by FP behavior evaluation computer code: FPRETAIN, which has been developed to understand the fuel behavior at extended burn-up stages. The birth, the release and the retain process of various kinds of fission products in the fuel pellet and fuel mechanical behavior were modelled in the code. In this work, open-to-public data obtained from several test reactors (Halden Boiling Water Reactor, Studsvik R2 Reactor and Riso DR-3 Reactor) and from the domestic commercial power reactor were utilized for verification. Concluding remarks obtained are: (1) Regarding fuel centerline temperatures as a function of rod linear power and burn-up, calculation coincided well with experiment up to the burn-ups range of 20 MWd/kgU. (2) At burn-ups within 20 MWd/kgU, increased plenum pressures by released Xe and Kr gases were well predicted. (3) On release rate of the stable gaseous FP to burn-up range up to 35 MWd/kgU, it was understood that the code was tended to make under-estimation, especially at increased burn-up stages. The code calculation against fuels from domestic reactors, however, resulted in good agreement with post irradiation data. (4) Within burn-up ranges up to 35 MWd/kgU, hoop strain calculated by the code coincided well with data obtained from experiment. (5) Regarding retained FP gas distribution in a fuel, the calculation agreed with experimental data qualitatively. (author)

  12. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    Victoria, M.; Baluc, N.; Spaetig, P.

    2001-01-01

    In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

  13. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  14. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

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

  15. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

  16. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

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

  17. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

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

  18. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  19. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  20. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  1. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  2. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

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

  3. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

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

  4. 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)

  5. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

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

  6. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

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

  7. The role of nuclear reactors in space exploration and development

    International Nuclear Information System (INIS)

    Lipinski, R.J.

    2000-01-01

    The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of 238 Pu for power and typically generate 235 U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. One reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built

  8. Boiling water reactors stability analysis - a challenge for the study of complex nonlinear dynamical systems

    International Nuclear Information System (INIS)

    Hennig, D.

    1997-01-01

    In boiling water reactors, there is a region in the operating map for which the reactor exhibits stable or unstable power oscillations. This oscillatory behaviour had to be understood in detail, in order to estimate, in a reliable way, the stability limits. This paper describes the BWR stability analysis methodology used at PSI and presents some recent results. (author) figs., tab., 38 refs

  9. New reactor concepts

    International Nuclear Information System (INIS)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

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

  10. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  11. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

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

  12. Multi purpose research reactor

    International Nuclear Information System (INIS)

    Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

    2006-01-01

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

  13. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  14. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

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

  15. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  16. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

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

  17. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

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

  18. Quality assurance in the manufacture of metallic uranium fuel for research reactors

    International Nuclear Information System (INIS)

    Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

    2009-01-01

    Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

  19. Stable isotopes in Lithuanian bioarcheological material

    Science.gov (United States)

    Skipityte, Raminta; Jankauskas, Rimantas; Remeikis, Vidmantas

    2015-04-01

    Investigation of bioarcheological material of ancient human populations allows us to understand the subsistence behavior associated with various adaptations to the environment. Feeding habits are essential to the survival and growth of ancient populations. Stable isotope analysis is accepted tool in paleodiet (Schutkowski et al, 1999) and paleoenvironmental (Zernitskaya et al, 2014) studies. However, stable isotopes can be useful not only in investigating human feeding habits but also in describing social and cultural structure of the past populations (Le Huray and Schutkowski, 2005). Only few stable isotope investigations have been performed before in Lithuanian region suggesting a quite uniform diet between males and females and protein intake from freshwater fish and animal protein. Previously, stable isotope analysis has only been used to study a Stone Age population however, more recently studies have been conducted on Iron Age and Late medieval samples (Jacobs et al, 2009). Anyway, there was a need for more precise examination. Stable isotope analysis were performed on human bone collagen and apatite samples in this study. Data represented various ages (from 5-7th cent. to 18th cent.). Stable carbon and nitrogen isotope analysis on medieval populations indicated that individuals in studied sites in Lithuania were almost exclusively consuming C3 plants, C3 fed terrestrial animals, and some freshwater resources. Current investigation demonstrated social differences between elites and country people and is promising in paleodietary and daily life reconstruction. Acknowledgement I thank prof. dr. G. Grupe, Director of the Anthropological and Palaeoanatomical State Collection in Munich for providing the opportunity to work in her laboratory. The part of this work was funded by DAAD. Antanaitis-Jacobs, Indre, et al. "Diet in early Lithuanian prehistory and the new stable isotope evidence." Archaeologia Baltica 12 (2009): 12-30. Le Huray, Jonathan D., and Holger

  20. Dynamical attraction to stable processes

    OpenAIRE

    Fisher, Albert M.; Talet, Marina

    2012-01-01

    We apply dynamical ideas within probability theory, proving an almost-sure invariance principle in log density for stable processes. The familiar scaling property (self-similarity) of the stable process has a stronger expression, that the scaling flow on Skorokhod path space is a Bernoulli flow. We prove that typical paths of a random walk with i.i.d. increments in the domain of attraction of a stable law can be paired with paths of a stable process so that, after applying a non-random regula...

  1. Reactor core monitor for nuclear reactor

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    The device of the present invention provides a various information of a wide adaptability, such as a power distribution, to an operator by determining a reactor core performance of the reactor by a performance calculation with improved accuracy. That is, a calculation means determines a neutron flux distribution of the reactor and coolant temperature based on the neutron flux distribution. A measuring means measures a cooled temperature of a reactor core inlet and a temperature at the exit of a fuel assembly. The result of coolant temperature by the measuring means and the result of the calculation by the calculation means are compared. The result of the calculation for the neutron flux distribution obtained by the calculation means is corrected based on the result of the comparison. The calculation means introduces calculation at higher accuracy by adopting two-dimensional balance in the fuel assembly. Further, a more accurate three-dimensional neutron diffusion calculation model is introduced in an on-line computer. Then, the accuracy of the calculation for the neutron flux distribution, power distribution, temperature distribution, etc. is improved. In view of the above, adaptability of a reactor core monitor is widened. (I.S.)

  2. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  3. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  4. Modeling of Reactor Kinetics and Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov

    2010-09-01

    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  5. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  6. Subchannel analysis in nuclear reactors

    International Nuclear Information System (INIS)

    Ninokata, H.; Aritomi, M.

    1992-01-01

    This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)

  7. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  8. A nuclear power reactor

    International Nuclear Information System (INIS)

    Borrman, B.E.; Broden, P.; Lundin, N.

    1979-12-01

    The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

  9. Reactor cost driving items

    International Nuclear Information System (INIS)

    Spears, W.R.

    1987-01-01

    Assuming that the design solutions presently perceived for NET can be extrapolated for use in a power reactor, and using costing experience with present day fusion experiments and with fission power plants, the major components of the cost of a tokamak fusion power reactor are described. The analysis shows the emphasis worth placing on various areas of plant design to reduce costs

  10. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

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

  11. CAREM 25 nuclear reactor

    International Nuclear Information System (INIS)

    Rossini, A.A.; Ordonez, J.P.; Rajoy, J.E.; Durione, C.

    1990-01-01

    This work describes the CAREM project reactor, its design philosophy, its main characteristics and its advantages with respect to similar reactors. The main objective is to use the nuclear energy at lower costs than those applied up to now. (Author) [es

  12. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

    Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

  13. International thermal reactor development

    International Nuclear Information System (INIS)

    Zebroski, E.L.

    1977-01-01

    The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities

  14. The fusion reactor

    International Nuclear Information System (INIS)

    Brennan, M.H.

    1974-01-01

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

  15. Advanced converter reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-01-01

    Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U 3 O 8 at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals

  16. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

  17. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  18. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Lineberry, M.J.

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

  19. Nuclear reactor instrumentation method

    International Nuclear Information System (INIS)

    Handa, Hiroyuki; Hayashi, Katsumi; Nemesawa, Shigeki; Nemoto, Yuji; Ohashi, Masahisa.

    1993-01-01

    The present invention can appropriately monitor the state of a reactor core in an FBR type reactor which has a system of storing spent fuel assemblies in a reactor container while reducing the weight and making the structure compact in the reactor. That is, a fuel assembly having a shield lacking portion in upper axial shields is disposed. The shield lacking portion defines neutrons' leaking path from the reactor core. The leakage of neutrons from the path is detected by a neutron monitor disposed just above the fuel assembly. With such a constitution, influence of neutrons from stored spent fuel assemblies disposed to the out side of the radial shields can be reduced by a shielding effect of the existent radial shields around the reactor core. Further, if a shield lacking portion is locally disposed in the region of the upper axial shields just below the neutron monitor, neutrons from the reactor core can be monitored while suppressing excessive neutron leakage. As a result, it is unnecessary to dispose shields on the outer side of the spent fuel assembly disposed in the reactor core. (I.S.)

  20. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  1. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  2. 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

  3. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

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

  4. Iris reactor conceptual design

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  5. Reactor control device

    International Nuclear Information System (INIS)

    Fukami, Haruo; Morimoto, Yoshinori.

    1981-01-01

    Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

  6. Mirror reactor surface study

    International Nuclear Information System (INIS)

    Hunt, A.L.; Damm, C.C.; Futch, A.H.; Hiskes, J.R.; Meisenheimer, R.G.; Moir, R.W.; Simonen, T.C.; Stallard, B.W.; Taylor, C.E.

    1976-01-01

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

  7. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  8. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  9. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

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

  10. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1984-08-01

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

  11. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

    1977-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1976-09-01

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

  13. Mirror reactor studies

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Bender, D.J.

    1977-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

  14. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

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

  15. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    International Nuclear Information System (INIS)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W.

    1994-06-01

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as 55 Fe, 60 Co, and 99 Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC's ''Technical Position on Waste Form,'' Revision 1

  16. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-06-01

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as {sup 55}Fe, {sup 60}Co, and {sup 99}Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC`s ``Technical Position on Waste Form,`` Revision 1.

  17. Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

    International Nuclear Information System (INIS)

    Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

    2003-01-01

    The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

  18. Iodine chemistry in a reactor regulation

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A. [Nuclear Regulatory Commission, Washington, DC (United States). Advisory Committee on Reactor Safeguards

    1996-12-01

    Radioactive iodine has always been an important consideration in the regulation of nuclear power reactors to assure the health and safety of the public. Regulators adopted conservatively bounding predictions of iodine behavior in the earliest days of the development of nuclear power because there was so little known about either accidents or the chemistry of iodine. Today there is a flood of new information and understanding of the chemistry of iodine under reactor accident conditions. This paper offers some thoughts on how the community of scientists engaged in the study of iodine chemistry can present the results of their work so that it is more immediately adopted by the regulator. It is suggested that the scientific community consider the concept of consensus standards so effectively used within the engineering community to define the status of the study of radioactive iodine chemistry for reactor safety. (author) 9 refs.

  19. Progress of PWR reactor fuels: OSIRIS equipments

    International Nuclear Information System (INIS)

    Colomez, G.; Farny, G.; Vidal, H.

    1981-09-01

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering [fr

  20. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  1. 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.

  2. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  3. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

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

  4. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

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

  5. Scram device for gas-cooled reactor

    International Nuclear Information System (INIS)

    Murakami, Atsushi; Takahashi, Suehiro.

    1989-01-01

    A scram device for gas-cooled reactors has a hopper disposed below a stand pipe standing upright passing through a reactor container and electromagnets disposed therein. It further comprises neutron absorbing steel balls maintained between the electromagnets and the hopper upon energization of the electromagnets. Upon emergency reactor shutdown, energization for the electromagnets is interrupted to drop the neutron absorption stainless steel balls into the reactor core. It is an object of the present invention to keep the mechanical strength of the electromagnets in a high temperature gas atmosphere and not to reduce the insulation performance. That is, coils for the electromagnets are constituted with a small oxide-insulated metal sheath cable (MI cable). As the feature of the MI cable, it can maintain the mechanical strength even when exposed to high temperature gas coolant and the insulation performance thereof does not reduce by virture of its gas sealing property. Accordingly, a scram device of stable reliability can be obtained. (K.M.)

  6. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    Hollis, A.A.; Mitchell, J.T.D.

    1977-12-01

    This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

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

    Science.gov (United States)

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

    2015-12-01

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

  8. Impact of closed Brayton cycle test results on gas cooled reactor operation and safety

    International Nuclear Information System (INIS)

    Wright, St.A.; Pickard, P.S.

    2007-01-01

    This report summarizes the measurements and model predictions for a series of tests supported by the U.S. Department of Energy that were performed using the recently constructed Sandia Brayton Loop (SBL-30). From the test results we have developed steady-state power operating curves, controls methodologies, and transient data for normal and off-normal behavior, such as loss of load events, and for decay heat removal conditions after shutdown. These tests and models show that because the turbomachinery operates off of the temperature difference (between the heat source and the heat sink), that the turbomachinery can continue to operate (off of sensible heat) for long periods of time without auxiliary power. For our test hardware, operations up to one hour have been observed. This effect can provide significant operations and safety benefits for nuclear reactors that are coupled to a Brayton cycles because the operating turbomachinery continues to provide cooling to the reactor. These capabilities mean that the decay-heat removal can be accommodated by properly managing the electrical power produced by the generator/alternator. In some conditions, it may even be possible to produce sufficient power to continue operating auxiliary systems including the waste heat circulatory system. In addition, the Brayton plant impacts the consequences of off-normal and accident events including loss of load and loss of on-site power. We have observed that for a loss of load or a loss of on-site power event, with a reactor scram, the transient consists initially of a turbomachinery speed increase to a new stable operating point. Because the turbomachinery is still spinning, the reactor is still being cooled provided the ultimate heat sink remains available. These highly desirable operational characteristics were observed in the Sandia Brayton loop. This type of behavior is also predicted by our models. Ultimately, these results provide the designers the opportunity to design gas

  9. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

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

  10. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  11. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  12. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

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

  13. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

  14. Water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Donaldson, A.J.

    1989-01-01

    In order to reduce any loss of primary water coolant from around a reactor core of a water cooled nuclear reactor caused by any failure of a pressure vessel, an inner vessel is positioned within and spaced from the pressure vessel. The reactor core and main portion of the primary water coolant circuit and a heat exchanger are positioned within the inner vessel to maintain some primary water coolant around the reactor core and to allow residual decay heat to be removed from the reactor core by the heat exchanger. In the embodiment shown an aperture at the upper region of the inner vessel is dimensioned configured and arranged to prevent steam from a steam space of an integral pressurised water cooled nuclear reactor for a ship entering the main portion of the primary water coolant circuit in the inner vessel if the longitudinal axis of the nuclear reactor is displaced from its normal substantially vertical position to an abnormal position at an angle to the vertical direction. Shields are integral with the inner vessel. (author)

  15. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  16. Reactor monitoring device

    International Nuclear Information System (INIS)

    Kono, Shigehiro.

    1996-01-01

    The device of the present invention monitors the stability of a power of a BWR type reactor by using each of recycling flow rates in addition to a reactor core flow rate to improve monitoring accuracy. Namely, a set value registering means is disposed for registering reactor core flow rate set values corresponding to the number of recycling flow rates not reaching a reference value for each of the recycling flow rates. A reactor flow rate take-out means judges whether each of the recycling flow rates reaches the reference value or not. The set values of the set value registering means are taken out based on the number of each of the recycling flow rate signals not reaching the reference values. The taken out set value and calculated reactor core flow rate value are compared by an abnormal alarm means. When calculated value is smaller than the set value, abnormality is informed. The accuracy for the monitoring is improved by monitoring the reactor power by using each of recycling flow rates in addition to the reactor core flow rate. (I.S.)

  17. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  18. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

    On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

  19. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  20. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

    Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)