WorldWideScience

Sample records for stable reactor behavior

  1. Improving Corrosion Behavior in SCWR, LFR and VHTR Reactor Materials by Formation of a Stable Oxide

    International Nuclear Information System (INIS)

    Motta, Arthur T.; Comstock, Robert; Li, Ning; Allen, Todd; Was, Gary

    2009-01-01

    The objective of this study is to understand the influence of the alloy microstructure and composition on the formation of a stable, protective oxide in the environments relevant to the SCWR and LFR reactor concepts, as well as to the VHTR. It is proposed to use state-of-the art techniques to study the fine structure of these oxides to identify the structural differences between stable and unstable oxide layers. The techniques to be used are microbeam synchrotron radiation diffraction and fluorescence, and cross-sectional transmission electron microcopy on samples prepared using focused ion beam.

  2. Moltex Energy's stable salt reactors

    International Nuclear Information System (INIS)

    O'Sullivan, R.; Laurie, J.

    2016-01-01

    A stable salt reactor is a molten salt reactor in which the molten fuel salt is contained in fuel rods. This concept was invented in 1951 and re-discovered and improved recently by Moltex Energy Company. The main advantage of using molten salt fuel is that the 2 problematic fission products cesium and iodine do not exist in gaseous form but rather in a form of a salt that present no danger in case of accident. Another advantage is the strongly negative temperature coefficient for reactivity which means the reactor self-regulates. The feasibility studies have been performed on a molten salt fuel composed of sodium chloride and plutonium/uranium/lanthanide/actinide trichloride. The coolant fluid is a mix of sodium and zirconium fluoride salts that will need low flow rates. The addition of 1 mol% of metal zirconium to the coolant fluid reduces the risk of corrosion with standard steels and the addition of 2% of hafnium reduces the neutron dose. The temperature of the coolant is expected to reach 650 Celsius degrees at the exit of the core. This reactor is designed to be modular and it will be able to burn actinides. (A.C.)

  3. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  4. Modelling aerosol behavior in reactor cooling systems

    International Nuclear Information System (INIS)

    McDonald, B.H.

    1990-01-01

    This paper presents an overview of some of the areas of concern in using computer codes to model fission-product aerosol behavior in the reactor cooling system (RCS) of a water-cooled nuclear reactor during a loss-of-coolant accident. The basic physical processes that require modelling include: fission product release and aerosol formation in the reactor core, aerosol transport and deposition in the reactor core and throughout the rest of the RCS, and the interaction between aerosol transport processes and the thermalhydraulics. In addition to these basic physical processes, chemical reactions can have a large influence on the nature of the aerosol and its behavior in the RCS. The focus is on the physics and the implications of numerical methods used in the computer codes to model aerosol behavior in the RCS

  5. 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

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

    International Nuclear Information System (INIS)

    Kukuda, Toshio

    1979-01-01

    The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov's method. For either parameter identification or adaptive control of a loosely-coupled-core reactor, there exists no canonical form of multiple input-multiple output system which can be directly applied for deriving the MRAS with the matrix version of the Kalman-Yakubovich lemma as it was in the case of the point reactor. This difficulty is circumvented by the practical assumption that the neutron density can be directly measured on each core as reactivity change is applied as input into the coupled core as a whole. For parameter identification, the model parameters are adaptively adjusted to those of each core, while for the adaptive control, plant parameters of each core can be adaptively compensated, again through control inputs, to asymptotically reduce the output error between the model and the plant. The point reactor is shown to correspond to a special case. (author)

  7. Dynamics and control of molten-salt breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sing, Vikram; Lish, Matthew R.; Chvala, Ondrej; Upadhyaya, Belle R. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    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.

  8. 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.

  9. Malfunction diagnosis and applications of stable adaptive schemes for a nuclear reactor system

    International Nuclear Information System (INIS)

    Fukuda, Toshio; Shibata, Heki.

    1979-01-01

    Malfunction diagnosis concerns a method to detect the abnormal phenomena during nuclear reactor operations, while stable adaptive schemes does the application of Model Reference Adaptive System (MRAS) to the nonlinear dynamics of a reactor for parameter identification and control. The new method for the malfunction diagnosis consists of the following ideas; an index defined as the sum of ratios of the square of a factor score to the contribution weight of the factor, which is evaluated by applying the multi-factor analysis technique to the data of the state of nuclear reactor systems like neutron flux, temperature, flow rate and so on. The excess of the index over some given threshold shows the reactor system would be in an abnormal state. Then a theory of optimal filtering by Kalman with the aid of the stochastic approximation is applied to estimate the neutron flux distribution at its abnormal state and subsequently the squared sum of difference between desirable and estimated flux distributions shows the spot at which the abnormal phenomena would have occurred in terms of the peak of its distribution. Parameter identification and adaptive control schemes are presented for a point reactor and a loosely-coupled-core reactor with internal feedbacks which lead to the nonlinearity of the overall system. Both schemes are shown stable with new representations of the systems, which correspond to the nonminimal system representation, in the vein of the 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. Some experiments of parameter identification for the thermal-hydraulic system are carried out successfully using a simplified channel in which flow rate is varied in a binary form. (J.P.N.)

  10. 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.

  11. Models of iodine behavior in reactor containments

    International Nuclear Information System (INIS)

    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

  12. 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

  13. Aerosol behavior and light water reactor source terms

    International Nuclear Information System (INIS)

    Abbey, F.; Schikarski, W.O.

    1988-01-01

    The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered

  14. 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)

  15. 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

  16. 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.

  17. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Bolme, A.B.

    1996-01-01

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  18. 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)

  19. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    Cullingford, H.S.; Edeskuty, F.J.

    1981-01-01

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  20. 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.)

  1. 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

  2. Sloshing behaviors in liquid metal reactors

    International Nuclear Information System (INIS)

    Choun, Young Sun; Yoo, Bong; Lee, Jae Han

    1998-04-01

    It is very difficult to predict sloshing response in the LMFBR which contains many internal components, since the sloshing response of a coolant depends not only upon the frequency contents of input ground motion, but also upon the configuration of reactor vessel and internal components, arrangement of internal components, and excitation direction of ground motion. Therefore, clear understanding of sloshing in the reactor vessel and development of proper technologies to predict sloshing response are necessary to establish the safety of the LMFBR against earthquake. The purpose of this report is to review sloshing behaviors in the LMFBR for the clear understanding of sloshing characteristics and to propose some interesting topics to be studied in future. (author). 34 refs., 12 tabs., 50 figs

  3. 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.

  4. 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.

  5. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  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. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon [Youngdong Univ., Yeongdong (Korea, Republic of)] (and others)

    2003-03-15

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study.

  8. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    International Nuclear Information System (INIS)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon

    2003-03-01

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study

  9. Silver-indium-cadmium control rod behavior and aerosol formation in severe reactor accidents

    International Nuclear Information System (INIS)

    Petti, D.A.

    1987-04-01

    Silver-indium-cadmium (Ag-In-Cd) control rod behavior and aerosol formation in severe reactor accidents are examined in an attempt to improve the methodology used to estimate reactor accident source terms. Control rod behavior in both in-pile and out-of-pile experiments is reviewed. A mechanistic model named VAPOR is developed that calculates the downward relocation and simultaneous vaporization behavior of the Ag-In-Cd alloy expected after control rod failure in a severe reactor accident. VAPOR is used to predict the release of silver, indium, and cadmium vapors expected in the Power Burst Facility (PBF) Severe Fuel Damage (SFD) 1-4 experiment. In addition, a sensitivity study is performed. Although cadmium is found to be the most volatile constituent of the alloy, all of the calculations predict that the rapid relocation of the alloy down to cooler portions of the core results in a small release for all three control rod alloy vapors. Potential aerosol formation mechanisms in a severe reactor accident are reviewed. Specifically, models for homogeneous, ion-induced, and heterogeneous nucleation are investigated. These models are applied to silver, cadmium, and CsI to examine the nucleation behavior of these three potential aerosol sources in a severe reactor accident and to illustrate the competition among these mechanisms for vapor depletion. The results indicate that aerosol formation in a severe reactor accident occurs in three stages. In the first stage, ion-induced nucleation causes aerosol generation. During the second stage, ion-induced and heterogeneous nucleation operates as competing pathways for gas-to-particle conversion until sufficient aerosol surface area is generated. In the third stage, ion-induced nucleation ceases; and heterogeneous nucleation becomes the dominant mechanism of gas-to-particle conversion until equilibrium is reached

  10. 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.

  11. 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

  12. 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

  13. Reactor thermal behaviors under kinetics parameters variations in fast reactivity insertion

    Energy Technology Data Exchange (ETDEWEB)

    Abou-El-Maaty, Talal [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)], E-mail: talal22969@yahoo.com; Abdelhady, Amr [Reactors Department, Atomic Energy Authority, Cairo 13759 (Egypt)

    2009-03-15

    The influences of variations in some of the kinetics parameters affecting the reactivity insertion are considered in this study, it has been accomplished in order to acquire knowledge about the role that kinetic parameters play in prompt critical transients from the safety point of view. The kinetics parameters variations are limited to the effective delayed neutron fraction ({beta}{sub eff}) and the prompt neutron generation time ({lambda}). The reactor thermal behaviors under the variations in effective delayed neutron fraction and prompt neutron generation time included, the reactor power, maximum fuel temperature, maximum clad temperature, maximum coolant temperature and the mass flux variations at the hot channel. The analysis is done for a typical swimming pool, plate type research reactor with low enriched uranium. The scram system is disabled during the accidents simulations. Calculations were done using PARET code. As a result of simulations, it is concluded that, the reactor (ETRR2) thermal behavior is considerably more sensitive to the variation in the effective delayed neutron fraction than to the variation in prompt neutron generation time and the fast reactivity insertion in both cases causes a flow expansion and contraction at the hot channel exit. The amplitude of the oscillated flow is a qualitatively increases with the decrease in both {beta}{sub eff} and {lambda}.

  14. 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 < 0.05), suggesting it is associated with longer-term 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

  15. 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.

    2016-01-01

    Background 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. The present 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. Methods 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 pre-resolution 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 pre-resolution year. Results 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 pre-resolution allocations to drinking and savings (OR = 1.77, 95% CI = 1.02 ∼ 3.08, p < .05), suggesting it is associated with longer term behavior regulation processes than abstinence. Conclusions The ASDE's unique predictive utility may rest on its comprehensive representation of contextual elements to support this patterning of

  16. Nuclear aerosol behavior during reactor accidents

    International Nuclear Information System (INIS)

    Williams, M.M.R.

    1990-01-01

    Some early reactor accidents are recalled together with their associated environmental consequences. One such consequence is the generation of radioactive aerosol. We described the various physical processes that such an aerosol cloud undergoes within the secondary containment building. These physical processes are then brought together quantitatively in a balance equation for the aerosol size spectrum as a function of position and time. Methods for solving this equation are discussed and illustrated by the method of moments based upon log-normal and modified gamma distributions. Current problems are outlined and directions for future work into aerosol behavior are suggested. (author)

  17. Uses of stable isotopes

    International Nuclear Information System (INIS)

    Axente, Damian

    1998-01-01

    The most important fields of stable isotope use with examples are presented. These are: 1. Isotope dilution analysis: trace analysis, measurements of volumes and masses; 2. Stable isotopes as tracers: transport phenomena, environmental studies, agricultural research, authentication of products and objects, archaeometry, studies of reaction mechanisms, structure and function determination of complex biological entities, studies of metabolism, breath test for diagnostic; 3. Isotope equilibrium effects: measurement of equilibrium effects, investigation of equilibrium conditions, mechanism of drug action, study of natural processes, water cycle, temperature measurements; 4. Stable isotope for advanced nuclear reactors: uranium nitride with 15 N as nuclear fuel, 157 Gd for reactor control. In spite of some difficulties of stable isotope use, particularly related to the analytical techniques, which are slow and expensive, the number of papers reporting on this subject is steadily growing as well as the number of scientific meetings organized by International Isotope Section and IAEA, Gordon Conferences, and regional meeting in Germany, France, etc. Stable isotope application development on large scale is determined by improving their production technologies as well as those of labeled compound and the analytical techniques. (author)

  18. Hydrogen behavior in light-water reactors

    International Nuclear Information System (INIS)

    Berman, M.; Cummings, J.C.

    1984-01-01

    The Three Mile Island accident resulted in the generation of an estimated 150 to 600 kg of hydrogen, some of which burned inside the containment building, causing a transient pressure rise of roughly 200 kPa (2 atm). With this accident as the immediate impetus and the improved safety of reactors as the long-term goal, the nuclear industry and the Nuclear Regulatory Commission initiated research programs to study hydrogen behavior and control during accidents at nuclear plants. Several fundamental questions and issues arise when the hydrogen problem for light-water-reactor plants is examined. These relate to four aspects of the problem: hydrogen production; hydrogen transport, release, and mixing; hydrogen combustion; and prevention or mitigation of hydrogen combustion. Although much has been accomplished, some unknowns and uncertainties still remain, for example, the rate of hydrogen production during a degraded-core or molten-core accident, the rate of hydrogen mixing, the effect of geometrical structures and scale on combustion, flame speeds, combustion completeness, and mitigation-scheme effectiveness. This article discusses the nature and extent of the hydrogen problem, the progress that has been made, and the important unresolved questions

  19. 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.

  20. 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.

  1. Inelastic Cyclic Deformation Behaviors of Type 316H Stainless Steel for Reactor Pressure Vessel of Sodium-Cooled Fast Reactor at Elevated Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji-Hyun; Hong, Seokmin; Koo, Gyeong-Hoi; Lee, Bong-Sang; Kim, Young-Chun

    2015-01-01

    Type 316H stainless steel is a primary candidate material for a reactor pressure vessel of a sodium-cooled fast (SFR) reactor which is under development in Korea. The reactor pressure vessel for a SFR is subjected to inelastic deformation induced by cyclic thermal stress. Fully reversed cyclic testing and ratcheting testing at elevated temperatures were performed to characterize the inelastic cyclic deformation behaviors of Type 316H stainless steel at the SFR operating temperature. It was found that cyclic hardening of Type 316H stainless steel was enhanced, and the accumulation of ratcheting deformation of Type 316H stainless steel was retarded at around the SFR operating temperature. The results of the tensile testing and the microstructural investigation for dislocated structures after the inelastic deformation testing showed that dynamic strain aging affected the inelastic cyclic deformation behavior of Type 316 stainless steel at around the SFR operating temperature.

  2. 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

  3. 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

  4. Identification of tertiary butyl alcohol (TBA)-utilizing organisms in BioGAC reactors using 13C-DNA stable isotope probing.

    Science.gov (United States)

    Aslett, Denise; Haas, Joseph; Hyman, Michael

    2011-09-01

    Biodegradation of the gasoline oxygenates methyl tertiary-butyl ether (MTBE) and ethyl tertiary-butyl ether (ETBE) can cause tertiary butyl alcohol (TBA) to accumulate in gasoline-impacted environments. One remediation option for TBA-contaminated groundwater involves oxygenated granulated activated carbon (GAC) reactors that have been self-inoculated by indigenous TBA-degrading microorganisms in ground water extracted from contaminated aquifers. Identification of these organisms is important for understanding the range of TBA-metabolizing organisms in nature and for determining whether self-inoculation of similar reactors is likely to occur at other sites. In this study (13)C-DNA-stable isotope probing (SIP) was used to identify TBA-utilizing organisms in samples of self-inoculated BioGAC reactors operated at sites in New York and California. Based on 16S rRNA nucleotide sequences, all TBA-utilizing organisms identified were members of the Burkholderiales order of the β-proteobacteria. Organisms similar to Cupriavidus and Methylibium were observed in both reactor samples while organisms similar to Polaromonas and Rhodoferax were unique to the reactor sample from New York. Organisms similar to Hydrogenophaga and Paucibacter strains were only detected in the reactor sample from California. We also analyzed our samples for the presence of several genes previously implicated in TBA oxidation by pure cultures of bacteria. Genes Mpe_B0532, B0541, B0555, and B0561 were all detected in (13)C-metagenomic DNA from both reactors and deduced amino acid sequences suggested these genes all encode highly conserved enzymes. One gene (Mpe_B0555) encodes a putative phthalate dioxygenase-like enzyme that may be particularly appropriate for determining the potential for TBA oxidation in contaminated environmental samples.

  5. Macroscopic behavior of fast reactor fuel subjected to simulated thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-06-01

    High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 μmoles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate

  6. Fuel Behavior Modeling Issues Associated with Future Fast Reactor Systems

    International Nuclear Information System (INIS)

    Yacout, A.M.; Hofman, G.L.; Lambert, J.D.B.; Kim, Y.S.

    2007-01-01

    Major issues of concern related to advanced fast reactor fuel behavior are discussed here with focus on phenomena that are encountered during irradiation of metallic fuel elements. Identification of those issues is part of an advanced fuel simulation effort that aims at improving fuel design and reducing reliance on conventional approach of design by experiment which is both time and resource consuming. (authors)

  7. 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

  8. In-reactor creep rupture behavior of the D9 alloys

    International Nuclear Information System (INIS)

    Puigh, R.J.; Hamilton, M.L.

    1986-06-01

    The uncertainties in the in-reactor stress rupture data have been significantly reduced with the acquisition of the Materials Open Test Assembly (MOTA) for testing of materials in the Fast Flux Test Facility (FFTF). The temperature uncertainty associated with irradiation in this vehicle is +- 5 0 C. Moreover, through the use of tag gases and an on-line cover gas monitoring system, on-line detection of specimen ruptures is possible during irradiation, thereby significantly reducing the uncertainty associated with the rupture times. Titanium additions, increases in nickel content and decreases in chromium content, which were made to improve the swelling response of 316 SS, resulted in an alloy class referred to as ''D9''. In-reactor stress rupture data from the MOTA experiment have been reported on two conditions of the D9-type alloys for exposure times corresponding to 2,400 hours at irradiation temperatures of 575, 605, 670, and 750 0 C. For these conditions the in-reactor rupture times were similar to those observed in thermal control tests. This report will describe both the in-reactor stress rupture behavior and the thermal control data for 20% cold work (CW) 316 SS and for 10 and 20% CW D9-type alloy over a similar temperature range for in-reactor exposure times corresponding to 13170 hr. and peak fast fluences corresponding to 17 x 10 22 n/cm 2 (E > 0.1 MeV)

  9. 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

  10. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  11. 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.

  12. Safety-oriented global analysis of reactor dynamics

    International Nuclear Information System (INIS)

    Belhadj, M.; Aldemir, T.

    1992-01-01

    It is well known that the asymptotic solutions of the non-linear systems encountered in reactor dynamics can change from stable to periodic or from periodic to chaotic with a very small change in system parameters and/or initial conditions. In that respect, determination of the domains of attraction (DOAs) in the state-space that contains the asymptotic solutions and the identification of the basins of attraction (BOAs) and lead to these DOAs usually requires a global analysis of reactor dynamics (as opposed to a local analysis through perturbation theory). From the standpoint of safety, the DOAs indicate whether the reactor behavior remains within the imposed constraints or not, and the BOAs show which initial conditions lead to safe operation. Due to the lack of a general theory, often the only feasible method for the global analysis of nonlinear systems is the direct integration of governing equations. However, direct integration can be computationally prohibitive, particularly if there is uncertainty on the values of the system parameters to be used in the analysis, and/or asymptotic system behavior is chaotic. In a recent study, a global analysis algorithm was presented to determine the structure of DOAs (and their probability distribution when there is uncertainty on the system parameters) more quickly than by direct integration. This paper shows how the new algorithm can be expanded to determine the BOAs of reactor dynamics equations as well as their DOAs

  13. Analysis of dynamic stability and safety of the reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    This document defines the approximations done for establishing a mathematical model of a reactor. Since the model should be used for safety analysis, it was important to choose a mathematical model less stable than the reactor itself. The analysis was performed on the analog computer RAS. Results obtained and conclusions concerned with three possible reactor accidents are presented [sr

  14. 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.

  15. 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

  16. Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors

    International Nuclear Information System (INIS)

    Chambers, R.; Laats, E.T.

    1981-01-01

    A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper

  17. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  18. 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.

  19. 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.)

  20. Ballooning stable high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki

    1981-04-01

    The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)

  1. Identification of process dynamics. Stability monitoring in BWR type reactors

    International Nuclear Information System (INIS)

    Abrahamsson, P.; Hallgren, P.

    1991-06-01

    Identification of process dynamics is used for stability monitoring in nuclear reactors (Boiling Water Reactor). This report treats the problem of estimating a damping factor and a resonance frequency from the neutron flux as measured in the reactor. A new parametric online method for identification is derived and presented, and is shown to meet the requirements of stability monitoring. The technique for estimating the process parameters is based on a recursive lattice filter algorithm. The problem of time varying parameters and offset, as well as offline experiments and signal processing are treated. All parts are implemented in a realtime program, using the language C. In comparison with earlier identifications, the new way of estimating the damping factor is shown to work well. Estimates of both the damping factor and the resonance frequency show a stable and reliable behavior. Future development and improvements are also indicated. (au)

  2. Large-scale experiments on aerosol behavior in light water reactor containments

    International Nuclear Information System (INIS)

    Schock, W.; Bunz, H.; Adams, R.E.; Tobias, M.L.; Rahn, F.J.

    1988-01-01

    Recently, three large-scale experimental programs were carried out dealing with the behavior of aerosols during core-melt accidents in light water reactors (LWRs). In the Nuclear Safety Pilot Plant (NSPP) program, the principal behaviors of different insoluble aerosols and of mixed aerosols were measured in dry air atmospheres and in condensing steam-air atmospheres contained in a 38-m/sup 3/ steel vessel. The Demonstration of Nuclear Aerosol Behavior (DEMONA) program used a 640-m/sup 3/ concrete containment model to simulate typical accident sequence conditions, and measured the behavior of different insoluble aerosols and mixed aerosols in condensing and transient atmospheric conditions. Part of the LWR Aerosol Containment Experiments (LACE) program was also devoted to aerosol behavior in containment; and 852-m/sup 3/ steel vessel was used, and the aerosols were composed of mixtures of insoluble and soluble species. The results of these experiments provide a suitable data base for validation of aerosol behavior codes. Fundamental insight into details of aerosol behavior in condensing environments has been gained through the results of the NSPP tests. Code comparisons have been and are being performed in the DEMONA and LACE experiments

  3. 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)

  4. Nolinear stability analysis of nuclear reactors : expansion methods for stability domains

    International Nuclear Information System (INIS)

    Yang, Chae Yong

    1992-02-01

    coefficient is positive, it is not necessarily true that an (linearly stable) operating power level is absolutely stable (unconditionally stable or stable in the large). In this situation, the stability region of the operating power level of the reactor is bounded and hence the reactor may experience sudden changes in power level for large perturbations. Furthermore, when mass flow rate decreases, the size of the stability region gradually decreases. The extent of stability (the stability region) of an operating state is dependent on the values of the reactor parameters. In the parameter space of the reactivity feedback coefficients, the reactor feedback system has three partitions: unconditionally stable (stable in the large), conditionally stable, and unstable regions. These partitions of course change as the state of the reactor, e.g., design and operating conditions, changes. It is desired that the reactor always operate in the unconditionally stable region; the reactor should be stable for any perturbation, by taking suitable values of the reactivity feedback coefficients. As a result of nonlinear stability analysis using the methods in this thesis, a desirable range of the reactivity feedback coefficients can be obtained for stable operation of a nuclear reactor

  5. 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)

  6. 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

  7. Study on light water reactor fuel behavior under reactivity initiated accident condition in TREAT

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Ishijima, Kiyomi; Ochiai, Masaaki; Tanzawa, Sadamitsu; Uemura, Mutsumi

    1981-05-01

    This report reviews the results of the fuel failure experiments performed in TREAT in the U.S.A. simulating Reactivity Initiated Accidents. One of the main purposes of the TREAT experiments is the study of the fuel failure behavior, and the other is the study of the molten fuel-water coolant interaction and the consequent hydrogen behavior. This report mainly shows the results of the TREAT experiments studying the fuel failure behavior in Light Water Reactor, and then it describes the fuel failure threshold and the fuel failure mechanism, considering the results of the photographic experiments of the fuel failure behavior with transparent capsules. (author)

  8. 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

  9. Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

    International Nuclear Information System (INIS)

    Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

    2014-01-01

    Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

  10. 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)

  11. Numerical simulation for debris bed behavior in sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Tagami, Hirotaka; Tobita, Yoshiharu

    2014-01-01

    For safety analysis of SFR, it is necessary to evaluate behavior along with coolability of debris bed in lower plenum which is formed in severe accident. In order to analyze debris behavior, model for dense sediment particles behavior was proposed and installed in SFR safety analysis code SIMMER. SIMMER code could adequately reproduce experimental results simulating the self-leveling phenomena with appropriate model parameters for bed stiffness. In reactor condition, the self-leveling experiment for prototypical debris bed has not been performed. Additionally, the prototypical debris bed consists of non-spherical particles and it is difficult to quantify model parameters. This situation brings sensitivity analysis to investigate effect of model parameters on the self-leveling phenomena of prototypical debris bed in present paper. As initial condition for sensitivity analysis, simple mound-like debris bed in sodium-filled lower plenum in reactor vessel is considered. The bed consists of the mixture of fuel debris of 3,300 kg and steel debris of 1,570 kg. Decay heat is given to this fuel debris. The model parameter is chosen as sensitivity parameter. Sensitivity analysis shows that the model parameters can effect on intensity of self-leveling phenomena and eventual flatness of bed. In all analyses, however, coolant and sodium vapor break the debris bed at mainly center part of bed and the debris is relocated to outside of bed. Through this process, the initial debris bed is almost planarized before re-melting of debris. This result shows that the model parameters affect the self-leveling phenomena, but its effect in the safety analysis of SFRs is limited. (author)

  12. Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo S, L F

    1991-12-31

    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).

  13. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantly affects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs

  14. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantlyaffects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs.

  15. 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

  16. 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)

  17. 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.

  18. 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.

  19. 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.)

  20. Behavior of antimony isotopes in the primary coolant of WWER-1000-type nuclear reactors in NPP Kozloduy during operation and shutdown

    International Nuclear Information System (INIS)

    Dobrevski, Ivan D.; Zaharieva, Neli N.; Minkova, Katia F.; Gerchev, Nikolay B.

    2009-01-01

    This paper focuses on the behavior of the antimony isotopes 122 Sb and 124 Sb in the coolant of the WWER reactors in the nuclear power plant Kozloduy (Bulgaria) during operation and shutdown. It is concluded that the chemical properties of their actual precursor, the isotope 121 Sb, determine the behavior of 122 Sb and 124 Sb during operation, load fluctuations, and shutdown as well as during the reactor coolant purification process. It is supposed that differences between the reactor bulk and the core fuel cladding surface chemistry as well as the presence of sub-cooled nucleate boiling at the fuel cladding may create conditions under which a local oxidizing environment may come into existence. (orig.)

  1. 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

  2. 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)

  3. The aqueous homogeneous suspension reactor project

    International Nuclear Information System (INIS)

    1975-01-01

    The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described

  4. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  5. Investigation on analytical method for the transfer behavior of corrosion product (CP) in the fast breeder reactors

    International Nuclear Information System (INIS)

    Matsuo, Youichirou; Sasaki, Shinji

    2013-05-01

    Radioactive corrosion products (CPs) are main cause of personal radiation exposure during maintenance work without fuel failure in FBR plants. The most important CP species are 54 Mn and 60 Co. The deposited radioactive CPs cause radiation fields near the piping and components, and the CPs contribute to the radiation exposure of the plant-worker. In this review, firstly, the collected knowledge about CP transfer behavior in the fast reactor are reviewed and analyzed. Secondly, the existing analytical methods to evaluate CP transfer behavior are investigated, issues of which and their solutions are extracted and discussed. Finally, examples of the calculated results by the improved analytical method are described. The provided conclusions are as follows; (1) Collected knowledge on CP transfer behavior. CP generation is mainly due to the dissolution of CP from hot reactor core constitution materials to hot sodium. On the core materials, particle-formed structure was confirmed. The evidence of CP precipitation in the low temperature part of the primary cooling system and the lower part of reactor core was provided. Similarly, the evidence of CP particle deposition in the same domain was also provided. (2) Extracted issues on analytical methods of CP transfer and proposed solutions. In the past, radioactivity caused by CP deposition on the piping and the core materials surface is confirmed. Subsequently, analytical models were developed based on the distribution of the CPs in the reactor coolant systems and the out-pile sodium loop test. The local high radiation dosage (such as elbow part) was observed by the radiation measurement. However, this behavior cannot be evaluated by the existing model, and it is considered necessary to take into account the transfer of CP particle. (3) The recent trend of the CP behavioral analysis method. Novel CP particle generation, transfer and deposition models were developed based on existing knowledge on CP behavior. The developed

  6. Data for FUMEX: Results from fuel behavior studies at the OECD Halden Reactor Project for model validation and development

    International Nuclear Information System (INIS)

    Wiesenack, W.

    1997-01-01

    Investigations of phenomena associated with extended or high burn-up are an important part of the fuel and materials testing programme carried out at the OECD Halden Reactor Project. The in-core studies comprise long term fuel rod behavior as well as the response to power ramps. Performance is assessed through measurements of fuel centre temperature, rod pressure, elongation of cladding and fuel stack, and cladding diameter changes obtained during full power reactor operation. Data from fuel behavior studies at the OECD Halden Reactor Project, provided for the IAEA co-ordinated research programme FUMEX, are used to elucidate short and long-term developments of fuel behavior. The examples comprise: fuel conductivity degradation manifested as a gradual temperature increase with burn-up; the influence of a combination of small gap/high fission gas release on fuel centre temperature (situation at high burn-up); fission gas release during normal operation and power ramps, and the possibility of a burn-up enhancement; PCMI reflected by cladding elongation, also for the case of a nominally open gap, and the change of interaction onset with burn-up. (author). 10 refs, 9 figs, 1 tab

  7. Data for FUMEX: Results from fuel behavior studies at the OECD Halden Reactor Project for model validation and development

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-08-01

    Investigations of phenomena associated with extended or high burn-up are an important part of the fuel and materials testing programme carried out at the OECD Halden Reactor Project. The in-core studies comprise long term fuel rod behavior as well as the response to power ramps. Performance is assessed through measurements of fuel centre temperature, rod pressure, elongation of cladding and fuel stack, and cladding diameter changes obtained during full power reactor operation. Data from fuel behavior studies at the OECD Halden Reactor Project, provided for the IAEA co-ordinated research programme FUMEX, are used to elucidate short and long-term developments of fuel behavior. The examples comprise: fuel conductivity degradation manifested as a gradual temperature increase with burn-up; the influence of a combination of small gap/high fission gas release on fuel centre temperature (situation at high burn-up); fission gas release during normal operation and power ramps, and the possibility of a burn-up enhancement; PCMI reflected by cladding elongation, also for the case of a nominally open gap, and the change of interaction onset with burn-up. (author). 10 refs, 9 figs, 1 tab.

  8. Thermohydraulic accident behavior of reactors

    International Nuclear Information System (INIS)

    Horche, W.; Kirmse, R.; Reichenbach, D.; Weber, J.P.

    1992-01-01

    GRS, on behalf of the German Federal Ministry for the Environment, conducted an assessment of the technical safety of the Greifswald nuclear generating units of the Soviet WWER-440/W-230 and W-213 reactor lines, respectively. The evaluation of existing accident analyses and the execution of some first calculations by GRS added to the know-how of GRS. This is reflected in the increased participation by GRS in international expert bodies investigating safety problems of WWER-440 plants. The contributions made towards international WWER projects within the framework of IAEA missions or as a result of bilateral consultations strengthen international partnership in the field of reactor safety in Central and Eastern Europe. (orig.) [de

  9. 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

  10. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    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.

  11. Unpredictably Stable

    DEFF Research Database (Denmark)

    Failla, Virgilio; Melillo, Francesca; Reichstein, Toke

    2014-01-01

    Is entrepreneurship a more stable career choice for high employment turnover individuals? We find that a transition to entrepreneurship induces a shift towards stayer behavior and identify job matching, job satisfaction and lock-in effects as main drivers. These findings have major implications...

  12. Fast reactor fuel pin behavior analyses in a LOF type transient event

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

    2013-06-01

    In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

  13. The CANDUR Reactor - The Practical Path to RU and TH use in Nuclear Reactors

    International Nuclear Information System (INIS)

    Kuran, Sermet; Yang, Dezi

    2012-01-01

    The CANDU heavy water reactor has unrivalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and Thorium (Th). Recently, this unique CANDU reactor feature attracted considerable attention due to favourable commercial, environmental and strategic needs. This paper summarizes the solid progress over the last three years and outlines CANDU Energy Incorporated's (CEI) multi-stage vision of utilizing various fuels in currently operational and new build CANDU reactors. In CEI's fuel-cycle vision, CANDU reactors will operate in conjunction with other reactor types and use advanced fuels to produce more energy and ensure the most efficient and least costly method of utilizing Light Water Reactor (LWR) used fuel. With this vision and the tandem goal of systematic adoption of Thorium based fuels, CANDU reactors will be a strong technology partner in ensuring the availability of long-term stable resources for nuclear power plants

  14. Structural analysis and modeling of water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Roshan Zamir, M.

    2000-01-01

    An important aspect of the design and analysis of nuclear reactor is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system under normal and emergency operating conditions. To achieve these objectives and in order to provide a suitable computer code based on fundamental material properties for design and study of the thermal-mechanical behavior of water reactor fuel rods during their irradiation life and also to demonstrate the fuel rod design and modeling for students, The KIANA-1 computer program has been developed by the writer at Amir-Kabir university of technology with support of Atomic Energy Organization of Iran. KIANA-1 is an integral one-dimensional computer program for the thermal and mechanical analysis in order to predict fuel rods performance and also parameter study of Zircaloy-clad UO 2 fuel rod during steady state conditions. The code has been designed for the following main objectives: To give a solution for the steady state heat conduction equation for fuel as a heat source and clad by using finite difference, control volume and semi-analytical methods in order to predict the temperature profile in the fuel and cladding. To predict the inner gas pressures due to the filling gases and released gaseous fission products. To predict the fission gas production and release by using a simple diffusion model based on the Booth models and an empirical model. To calculate the fuel-clad gap conductance for cracked fuel with partial contact zones to a closed gap with strong contact. To predict the distribution of stress in three principal directions in the fuel and sheet by assuming one-dimensional plane strain and asymmetric idealization. To calculate the strain distribution in three principal directions and the corresponding deformation in the fuel and cladding. For this purpose the permanent strain such as creep or plasticity as well as the thermoelastic deformation and also the swelling, densification, cracking

  15. Effects of irradiation and thermal aging upon fatigue-crack growth behavior of reactor pressure boundary materials. [Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    James, L. A.

    1978-10-01

    Two processes that have the potential to produce degradation in the properties of pressure boundary materials are neutron irradiation and long-time thermal aging. This paper uses linear-elastic fracture mechanics techniques to assess the effect of these two processes upon the fatigue-crack growth behavior of a number of alloys commonly employed in reactor pressure boundaries. The materials evaluated include ferritic steels, austenitic stainless steels, and nickel-base alloys typical of those employed in a number of reactor types including water-cooled, gas-cooled, and liquid-metal-cooled designs.

  16. 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.)

  17. 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

  18. Investigation on the radiation damage behavior of various alloys in a fusion reactor using thorium molten salt

    International Nuclear Information System (INIS)

    Ubeyli, Mustafa; Demir, Teyfik

    2008-01-01

    In fusion reactors, one of the most important problems is the need for the frequent change of the first wall material during the reactor's operation due to the radiation damage induced by high energetic particles, especially fusion neutrons coming from fusion plasma. In order to solve this problem, in HYLIFE-II fusion reactor design, a liquid wall between the fusion plasma and first wall is used. This study presents the radiation damage behaviors of candidate structural materials (9Cr-2WVTa, V-4Cr-4Ti and W-5Re alloys) considered to be used in fusion reactors to determine the optimum thickness of the liquid wall in HYLIFE-II fusion reactor. In the liquid wall, a thorium molten salt consisting of 75%LiF-23%ThF 4 -2% 233 UF 4 was used. Calculations were carried out with respect to the variable liquid wall thickness and for an operation period of 30 years. Numerical results related to atomic displacement and helium generation damage pointed out that the liquid wall thickness should be at least 42, 66 and 81 cm for the materials, W-5Re, 9Cr-2WVTa, V-4Cr-4Ti, respectively in order not to exceed relevant damage limits after a reactor operation of 30 years

  19. 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

  20. 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

  1. 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

  2. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  3. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  4. Startup measurements on the CABRI reactor

    International Nuclear Information System (INIS)

    Kussmaul, G.; Bensoussan, P.; Dadillon, J.; Golinelli, C.; Tonolli, J.

    1979-08-01

    The CABRI reactor will be used for the investigation of the behavior of fresh and irradiated fast reactor fuel pins under TOP conditions. A startup programme has been carried out to measure fundamental data determining steady state and transient behavior of the driver core as well as data ensuring safe operation of the reactor. Special emphasis was laid on quantities not well known from previous neutronics calculations, e.g. prompt-neutron generation time, Doppler feedback and time-dependent reactivity injection. Utilizing the data inferred from measurements in the dynamic code DULCINEE good agreement between calculated and observed transient behavior of the driver core has been found

  5. Pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    1992-01-01

    The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude

  6. Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

    2015-01-01

    Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

  7. 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

  8. Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

    International Nuclear Information System (INIS)

    Feng, K.M.; Huang, J.H.

    1994-01-01

    The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

  9. The behavior of radioactive iodine at the time of reactor accident and its counterplan

    International Nuclear Information System (INIS)

    Murata, Toshifumi

    1974-01-01

    When an accident occurs in a reactor, very volatile radioactive iodine is most dangerous among fission products. Among the isotopes of radioactive iodine, 131 I which has longer half-life is harmful. Supposing one-tenth of the radioactivity of 10 8 Ci in a reactor of 10 6 Kw heat output is due to 131 I, it weights about 100g. The behavior of the radioactive iodine is greatly subjected to the influence of inside temperature and other conditions. Therefore, very prudent policies are adopted by installing emergency core cooling system, containment vessels, and activated carbon filters. For the emergency core cooling system, water spraying, flooding with low pressure water, and maintaining of water level by high pressure water injection are adopted, while in the containment vessels, measures are taken so as to lower the inside pressure and minimize leakage. (Kobatake, H.)

  10. 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

  11. 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

  12. 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

  13. 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)

  14. Assessment of models for steam release from concrete and implications for modeling corium behavior in reactor cavities

    International Nuclear Information System (INIS)

    Washington, K.E.; Carroll, D.E.

    1988-01-01

    Models for concrete outgassing have been developed and incorporated into a developmental version of the CONTAIN code for the assessment of corium behavior in reactor cavities. The resultant code, referred to as CONTAIN/OR in order to distinguish it from the released version of CONTAIN, has the capability to model transient heat conduction and concrete outgassing in core-concrete interaction problems. This study focused on validation and assessment of the outgassing model through comparisons with other concrete response codes. In general, the model is not mechanistic; however, there are certain important processes and feedback effects that are treated rigorously. The CONTAIN outgassing model was compared against two mechanistic concrete response codes (USINT and SLAM). Gas release and temperature profile predictions for several concrete thicknesses and heating rates were performed with acceptable agreement seen in each case. The model was also applied to predict corium behavior in a reactor cavity for a hypothetical severe accident scenario. In this calculation, gases evolving from the concrete during nonablating periods fueled exothermic Zr chemical reactions in the corium. Higher corium temperatures and more concrete ablation were observed when compared with that seen when concrete outgassing was neglected. Even though this result depends somewhat upon the makeup of the corium sources and the concrete type in the cavity, it does show that concrete outgassing can be important in the modeling of corium behavior in reactor cavities. In particular, the need to expand the traditional role of CORCON from steady-state ablation to the consideration of more transient events is clearly evident as a result of this work. 5 refs., 11 figs., 1 tab

  15. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-08-01

    Results from the previously conducted Semiscale Mod-1 ECC injection test series were analyzed. Testing in the LOFT counterpart test series was essentially completed, and the steam generator tube rupture test series was begun. Two tests in the alternate ECC injection test series were conducted which included injection of emergency core coolant into the upper plenum through use of the low pressure injection system. The Loss-of-Fluid Test Program successfully completed nonnuclear Loss-of-Coolant Experiment L1-4. A nuclear test, GC 2-3, in the Power Burst Facility Reactor was performed to evaluate the power oscillation method of determining gap conductance and to determine the effects of initial gap size, fill gas composition, and fuel density on the thermal performance of a light water reactor fuel rod. Additional test results were obtained relative to the behavior of irradiated fuel rods during a fast power increase and during a high power film boiling transient. Fuel model development and verification activities continued for the steady state and transient Fuel Rod Analysis Program, FRAP-S and FRAP-T. A computer code known as RELAP4/MOD7 is being developed to provide best-estimate modeling for reflood during a postulated loss-of-coolant accident (LOCA). A prediction of the fourth test in the boiling water reactor (BWR) Blowdown/Emergency Core Cooling Program was completed and an uncertainty analysis was completed of experimental steady state stable film boiling data for water flowing vertically upward in round tubes. A new multinational cooperative program to study the behavior of entrained liquid in the upper plenum and cross flow in the core during the reflood phase of a pressurized water reactor LOCA was defined.

  16. 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.)

  17. The behavior of radioactive 137Cs and stable Cs at the isolated undisturbed mountain pond in Fukui, Japan

    International Nuclear Information System (INIS)

    Iwamoto, Kazumi; Kimura, Makio; Ando, Kenji; Amano, Hikaru

    2003-01-01

    The behavior of radioactive 137 Cs and stable Cs at the isolated undisturbed mountain pond in Fukui, Japan was studied for the pond water, the sedimentary grains and the soil near the pond. The concentrations of 137 Cs and stable Cs in the pond water ranged from 0.23 to 0.85 Bq/m 3 and from 0.005 to 0.018 mg/m 3 , respectively. The sedimentary grains were sorted by sieving into fractions with diameter from 2 mm to less than 38 μm. The concentrations of 137 Cs and stable Cs in the sorted grains were measured, and those of the adsorbed state were determined by subtracting the concentration of the grain matrix. The adsorbed concentrations increased with decrease in particle diameter and depended less on the kind of samples. The in-situ distribution coefficient Kd depended largely on particle diameter and increased with the decrease in diameter. The values of Kd ranged from about 20 to 1200 m 3 /kg for stable Cs and about 15 to 1000 m 3 /kg for 137 Cs, and the Kd of 137 Cs seemed to be slightly smaller than that of stable Cs. The concentration of stable Cs in the sedimentary mud was found to be close to that of the fine grains. The concentrations of stable Cs in the soil near the pond was about 7.7 mg/kg, and that of 137 Cs was about 0.6 kBq/kg for the surface layer soil and decreased with increase in soil depth. (author)

  18. 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)

  19. 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.

  20. 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

  1. Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Mori, Takero

    2005-01-01

    In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

  2. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  3. Tempered stable laws as random walk limits

    OpenAIRE

    Chakrabarty, Arijit; Meerschaert, Mark M.

    2010-01-01

    Stable laws can be tempered by modifying the L\\'evy measure to cool the probability of large jumps. Tempered stable laws retain their signature power law behavior at infinity, and infinite divisibility. This paper develops random walk models that converge to a tempered stable law under a triangular array scheme. Since tempered stable laws and processes are useful in statistical physics, these random walk models can provide a basic physical model for the underlying physical phenomena.

  4. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

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

    International Nuclear Information System (INIS)

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

    1990-06-01

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

  6. 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)

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

    International Nuclear Information System (INIS)

    Frid, W.

    1988-04-01

    Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. 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 that are addressed in this work

  8. Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

    2004-01-01

    Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

  9. 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

  10. Fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Vold, E.L.; Conn, R.W.

    1986-01-01

    Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

  11. Tellurium behavior during and after the TMI-2 accident

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Osetek, D.J.; Hobbins, R.R.

    1984-01-01

    The estimated behavior of tellurium during and after the accident at the Three Mile Island Unit-2 is presented. The behavior is based on all available measurement data for /sup 129m/Te, 132 Te and stable tellurium ( 126 Te, 128 Te and 130 Te), and best estimate calculations of tellurium release and transport. The predicted release was calculated using current techniques that relate release rate to fuel temperature and holdup of tellurium in zircaloy until significant oxidation occurs. The calculated release fraction was low, approximately 7%, but the total measured release for samples analyzed to date is about 4.0%. Of the measured tellurium about 2.4, 0.88, 0.42, 0.17 and 0.086% of core inventory were in the containment sump water, containment solids in water, makeup and purification demineralizer, containment inside surface, and the reactor primary coolant, respectively. A significant fraction (54%) of the calculated tellurium retained on the upper plenum surfaces (4.61% of the core inventory) was deposited during the high pressure injection of coolant at about 200 minutes after the reactor scram. Comparison of tellurium behavior with inpile and out-of-pile tests strongly suggests that zircaloy holds tellurium until significant cladding oxidation occurs

  12. 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

  13. Study on dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment

    International Nuclear Information System (INIS)

    Abe, K.; Kohyama, A.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    2001-01-01

    A Japan-USA Program of irradiation experiments for fusion research, 'JUPITER', has been established as a 6 year program from 1995 to 2000. The goal is to study the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment using fission reactors. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. The irradiation capsules for in-situ measurement and varying temperature were developed successfully. It was found that insulating ceramics were worked up to 3 dpa. The property changes and related issues in low activation structural materials were summarized. (author)

  14. Fusion reactor materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

  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. dynamic performance of research reactors

    International Nuclear Information System (INIS)

    Abo elnor, A.G.M.

    2007-01-01

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

  17. Non-traditional stable isotope behaviors in immiscible silica-melts in a mafic magma chamber.

    Science.gov (United States)

    Zhu, Dan; Bao, Huiming; Liu, Yun

    2015-12-01

    Non-traditional stable isotopes have increasingly been applied to studies of igneous processes including planetary differentiation. Equilibrium isotope fractionation of these elements in silicates is expected to be negligible at magmatic temperatures (δ(57)Fe difference often less than 0.2 per mil). However, an increasing number of data has revealed a puzzling observation, e.g., the δ(57)Fe for silicic magmas ranges from 0‰ up to 0.6‰, with the most positive δ(57)Fe almost exclusively found in A-type granitoids. Several interpretations have been proposed by different research groups, but these have so far failed to explain some aspects of the observations. Here we propose a dynamic, diffusion-induced isotope fractionation model that assumes Si-melts are growing and ascending immiscibly in a Fe-rich bulk magma chamber. Our model offers predictions on the behavior of non-traditional stable isotope such as Fe, Mg, Si, and Li that are consistent with observations from many A-type granitoids, especially those associated with layered intrusions. Diffusion-induced isotope fractionation may be more commonly preserved in magmatic rocks than was originally predicted.

  18. 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.

  19. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    Kempf, Rodolfo; Fortis, Ana M.

    2007-01-01

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

  20. 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.)

  1. The effects of phosphorus and boron on the behavior of a titanium-stabilized austenitic stainless steel developed for fast reactor service

    International Nuclear Information System (INIS)

    Hamilton, M.L.; Johnson, G.D.; Puigh, R.J.; Garner, F.A.; Maziasz, P.J.; Yang, W.J.S.; Abraham, N.

    1988-08-01

    Austenitic stainless steels are used for core component materials in liquid metal cooled reactors (LMRs). To extend the lifetime of LMR fuel assemblies, considerable effort was expended by the US breeder materials program to find ways to minimize radiation-induced dimensional changes (swelling and creep) and to maximize the creep rupture strength. After various elements were shown to strongly affect swelling and creep behavior, compositional modifications to a commercial grade austenitic stainless steel (AISI 316) produced an alloy with significant improvement in swelling resistance over the standard 300 series alloys. Changes were primarily in the concentrations of chromium, nickel, silicon and titanium, ASTM specification A771-83 was approved in 1983 for the new alloy, designated UNS S38660. Substantial improvement can be produced in the creep rupture behavior of this alloy. Elements such as phosphorus and boron, typically present in trace quantities, have a significant influence on the creep strength of austenitic stainless steels. Several heats of alloy S38660 were made that systematically varied the phosphorus and boron contents. Uniaxial creep tests were conducted at 704/degree/C (1300/degree/F) to evaluate the effects of these elements on the creep rate and the rupture life. The results of these tests were used to guide the production of reactor grade fuel pin cladding for further evaluations. Pressurized tube specimens were tested in the laboratory and also in a fast reactor. Results of these investigations have shown that the elements phosphorus and boron, present in minute but controlled amounts, increase both the in- reactor and ex-reactor rupture life and reduce both in-reactor swelling and creep rate. Microstructural evaluations were also conducted to help ascertain the mechanisms by which the improved properties were obtained. 41 refs., 28 figs., 3 tabs

  2. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  3. 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.

  4. Application of the reactor kinetics equations to the reactor safety analysis

    International Nuclear Information System (INIS)

    Sdouz, G.

    1976-01-01

    The reactor kinetics equations which can be solved by the computer program AIREK-III are used to describe the behavior of fast reactivity transients. By supplementing this computer program it was possible to solve additional safety problems, e.g. the course of reactor excursions induced by any form of reactivity input, the control of reactivity input as a function of a threshold-energy and the computation of produced energy. (author)

  5. Nonlinear behavior of multiple-helicity resistive interchange modes near marginally stable states

    International Nuclear Information System (INIS)

    Sugama, Hideo; Nakajima, Noriyoshi; Wakatani, Masahiro.

    1991-05-01

    Nonlinear behavior of resistive interchange modes near marginally stable states is theoretically studied under the multiple-helicity condition. Reduced fluid equations in the sheared slab configuration are used in order to treat a local transport problem. With the use of the invariance property of local reduced fluid model equations under a transformation between the modes with different rational surfaces, weakly nonlinear theories for single-helicity modes by Hamaguchi and Nakajima are extended to the multiple-helicity case and applied to the resistive interchange modes. We derive the nonlinear amplitude equations of the multiple-helicity modes, from which the convective transport in the saturated state is obtained. It is shown how the convective transport is enhanced by nonlinear interaction between modes with different rational surfaces compared with the single-helicity case. We confirm that theoretical results are in good agreement with direct numerical simulations. (author)

  6. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    International Nuclear Information System (INIS)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T

  7. Reactor kinetics - pulse and steady state

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B F; Morris, F M [Sandia Laboratories (United States)

    1974-07-01

    An analytical model has been developed which couples the nuclear and thermal characteristics of the Annular Core Pulse Reactor (ACPR) into a solution which describes both the neutron kinetics of the reactor and the temperature behavior of a fuel-moderator element. The model describes both pulse and steady state operations. This paper describes the important aspects of the reactor, the fuel- moderator elements, the neutron kinetic equations of the reactor, and the time-temperature behavior of a fuel-moderator element that is being subjected to the maximum power density in the core. The parameters which are utilized in the equations are divided into two classes, those that can be measured directly and those that are assumed to be known (each is described briefly). Some of the solutions which demonstrate the versatility of the analytical model are described. (author)

  8. 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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

    International Nuclear Information System (INIS)

    Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

    2012-01-01

    Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

  11. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

    International Nuclear Information System (INIS)

    Mathieu, L.

    2005-09-01

    Producing nuclear energy in order to reduce the anthropic CO 2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  12. Co-pyrolysis behavior of fermentation residues with woody sawdust by thermogravimetric analysis and a vacuum reactor.

    Science.gov (United States)

    Zheng, Yan; Zhang, Yimin; Xu, Jingna; Li, Xiayang; Charles Xu, Chunbao

    2017-12-01

    This study aimed at cost-effective utilization of fermentation residues (FR) from biogas project for bio-energy via co-pyrolysis of FR and woody sawdust (WS). In this study, a vacuum reactor was used to study the pyrolysis behaviors of individual and blend samples of FR and WS. Obvious synergistic effects were observed, resulting in a lower char yield but a higher gas yield. The presence of woody sawdust promoted the devolatilization of FR, and improved the syngas (H 2 and CO) content in the gaseous products. Compared to those of the char from pyrolysis of individual feedstock, co-pyrolysis of FR and WS in the vacuum reactor promoted the cracking reactions of large aromatic rings, enlarged the surface area and reduced the oxygenated groups of the resulted char. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Approach to securing of stable nuclear fuel supplies

    International Nuclear Information System (INIS)

    Koike, Kunihisa; Imamura, Isao; Noda, Tetsuya

    2010-01-01

    With the dual objectives of not only ensuring stable electric power supplies but also preventing global warming, the construction of new nuclear power plants is being planned in many countries throughout the world. Toshiba and Westinghouse Electric Company (WEC), a member of the Toshiba Group, are capable of supplying both boiling water reactor (BWR) and pressurized water reactor (PWR) plants to satisfy a broad range of customer requirements. Furthermore, to meet the growing demand for the securing of nuclear fuel supplies, Toshiba and WEC have been promoting the strengthening and further expansion of supply chains in the fields of uranium production, uranium hexafluoride (UF 6 ) conversion, uranium enrichment, and fuel fabrication. (author)

  14. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

    2014-01-01

    A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

  15. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  16. 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)

  17. 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)

  18. 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

  19. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  20. 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

  1. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  2. 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.)

  3. 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

  4. Development of a simultaneous partial nitrification and anaerobic ammonia oxidation process in a single reactor.

    Science.gov (United States)

    Cho, Sunja; Fujii, Naoki; Lee, Taeho; Okabe, Satoshi

    2011-01-01

    Up-flow oxygen-controlled biofilm reactors equipped with a non-woven fabric support were used as a single reactor system for autotrophic nitrogen removal based on a combined partial nitrification and anaerobic ammonium oxidation (anammox) reaction. The up-flow biofilm reactors were initiated as either a partial nitrifying reactor or an anammox reactor, respectively, and simultaneous partial nitrification and anammox was established by careful control of the aeration rate. The combined partial nitrification and anammox reaction was successfully developed in both biofilm reactors without additional biomass inoculation. The reactor initiated as the anammox reactor gave a slightly higher and more stable mean nitrogen removal rate of 0.35 (±0.19) kg-N m(-3) d(-1) than the reactor initiated as the partial nitrifying reactor (0.23 (±0.16) kg-N m(-3) d(-1)). FISH analysis revealed that the biofilm in the reactor started as the anammox reactor were composed of anammox bacteria located in inner anoxic layers that were surrounded by surface aerobic AOB layers, whereas AOB and anammox bacteria were mixed without a distinguishable niche in the biofilm in the reactor started as the partial nitrifying reactor. However, it was difficult to efficiently maintain the stable partial nitrification owing to inefficient aeration in the reactor, which is a key to development of the combined partial nitrification and anammox reaction in a single biofilm reactor. Copyright © 2010 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

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

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

    OpenAIRE

    Yogi W. Budhi; M. Effendy; Yazid Bindar; Subagjo

    2014-01-01

    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 ext...

  7. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  8. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    International Nuclear Information System (INIS)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-01-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

  9. 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.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    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 (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  12. 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

  13. 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

  14. The in-reactor deformation of the PCA alloy

    International Nuclear Information System (INIS)

    Puigh, R.J.

    1986-04-01

    The swelling and in-reactor creep behaviors of the PCA alloy have been determined from the irradiation of pressurized tube specimens in the FFTF reactor. These data have been obtained to a peak neutron fluence corresponding to approximately 80 dpa in the FFTF reactor for irradiation temperatures between 400 and 750 0 C. Diametral measurements performed on the unstressed specimens indicate the possible onset of swelling in the PCA alloy for irradiation temperatures between 400 and 550 0 C and at a neutron fluence corresponding to ∼50 dpa. The creep data suggest a non-linear fluence dependence and linear stress dependence (for hoop stresses less than 100 MPa) which is consistent with the in-reactor creep behavior of many cold worked austenitic stainless steels. These PCA creep data are compared to available 316 SS in-reactor creep data. The in-reactor creep strains for PCA are significantly less than observed in 20% cold worked 316 SS over the temperature ranges and fluences investigated

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

    International Nuclear Information System (INIS)

    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: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. 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

  16. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

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

  17. Molten salt reactors - safety options galore

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1997-01-01

    Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

  18. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T.

  19. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

    2014-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  20. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

    2016-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  1. Determining neutron multiplication factor in the infinite system by reactivity dependence on one dimension of the reactor core

    International Nuclear Information System (INIS)

    Pesic, M.

    1975-01-01

    The objective of this task was to apply Fermi age theory for determining τ and neutron multiplication factor in infinite medium by measuring reactivity coefficient of heavy water in heterogeneous mixed reactor lattice. Basis of experiment is the measurement of stable reactor period. Measurement of heavy water reactivity coefficient by measuring the stable reactor period is described for chosen overcritical heavy water levels. Calculated values of infinite multiplication factor for measured neutron age data are presented and they are compared to expected theoretical values

  2. Pulsed Compression Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roestenberg, T. [University of Twente, Enschede (Netherlands)

    2012-06-07

    The advantages of the Pulsed Compression Reactor (PCR) over the internal combustion engine-type chemical reactors are briefly discussed. Over the last four years a project concerning the fundamentals of the PCR technology has been performed by the University of Twente, Enschede, Netherlands. In order to assess the feasibility of the application of the PCR principle for the conversion methane to syngas, several fundamental questions needed to be answered. Two important questions that relate to the applicability of the PCR for any process are: how large is the heat transfer rate from a rapidly compressed and expanded volume of gas, and how does this heat transfer rate compare to energy contained in the compressed gas? And: can stable operation with a completely free piston as it is intended with the PCR be achieved?.

  3. 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)

  4. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  5. Micromechanisms of ductile stable crack growth in nuclear pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Belcher, W.P.A.; Druce, S.G.

    1981-10-01

    The objective of this work was to investigate the relationship between the micromechanisms of ductile crack growth, the microstructural constituent phases present in nuclear pressure vessel steel, and the observed fracture behavior as determined by impact and fracture mechanics tests. Results from a microstructural and mechanical property comparison of an A508 Class 3 pressurized water reactor nozzle forging cutout and a 150-mm-thick A533B Class 1 plate are reported. The variation of upper-shelf toughness between the two steels and its orientation sensitivity are discussed on the basis of inclusion and precipitate distributions. Inclusion clusters in A533B, deformed to elongated disks in the rolling plane, have a profound effect on short transverse fracture properties. Data derived using the multi-specimen J-integral method to characterize the initiation of ductile crack extension and resistance to stable crack growth are compared with equivalent Charpy results. Results of the J /SUB R/ -curve analyses indicate (1) that the A533B short transverse crack growth resistance is approximately half that observed from transverse and longitudinal specimen orientations, and (2) that the A508 initiation toughness and resistance to stable crack growth are insensitive to position through the forging wall, and are higher than exhibited by A533B at any orientation in the midthickness position.

  6. 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

  7. An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Maeda, Shigetaka; Mitsugashira, Toshiaki

    2005-01-01

    Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the 241 Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the 240 Pu to 239 Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor

  8. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  9. Effect of moving distance of temperature distribution on thermal ratchetting behavior of a FBR reactor vessel

    International Nuclear Information System (INIS)

    Ueta, Masahiro; Douzaki, Kouji; Takahashi, Yukio; Ooka, Yuji; Osaki, Toshio; Take, Kouji.

    1992-01-01

    It should be considered in a FBR reactor vessel design that thermal ratchetting might be caused by moving axial thermal gradient, in other words, moving sodium level. The behavior and the mechanism of ratchetting have almost become clear by studies for the past several years. A simplified evaluation method for ratchetting behavior has been proposed. However, the evaluation method has been shown to be excessively conservative by testing results. In this paper, the effect of moving distance of axial temperature distributions, which is one of main factors to be considered in precise estimation of ratchetting behavior, is studied by inelastic analyses. Based on the study, it is proposed to introduce a strain reducing factor taking account of residual stresses in the region of moving axial temperature distribution to the original evaluation method. The new method has been validated by comparing the prediction with results of both testing and the original method. (author)

  10. 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.

  11. 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)

  12. 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.)

  13. 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)

  14. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    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)

  15. 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)

  16. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  17. 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

  18. 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.)

  19. Naval application of battery optimized reactor integral system

    International Nuclear Information System (INIS)

    Kim, N. H.; Kim, T. W.; Son, H. M.; Suh, K. Y.

    2007-01-01

    Past civilian N.S. Savanna (80 MW t h), Otto-Hahn (38 MW t h) and Mutsu (36 MW t h) experienced stable operations under various sea conditions to prove that the reactors were stable and suitable for ship power source. Russian nuclear icebreakers such as Lenin (90 MW t h x2), Arukuchika (150 MW t h x2) showed stable operations under severe conditions during navigation on the Arctic Sea. These reactor systems, however, should be made even more efficient, compact, safe and long life, because adding support from the land may not be available on the sea. In order to meet these requirements, a compact, simple, safe and innovative integral system named Naval Application Vessel Integral System (NAVIS) is being designed with such novel concepts as a primary liquid metal coolant, a secondary supercritical carbon dioxide (SCO 2 ) coolant, emergency reactor cooling system, safety containment and so on. NAVIS is powered by Battery Optimized Reactor Integral System (BORIS). An ultra-small, ultra-long-life, versatile-purpose, fast-spectrum reactor named BORIS is being developed for a multi-purpose application such as naval power source, electric power generation in remote areas, seawater desalination, and district heating. NAVIS aims to satisfy special environment on the sea with BORIS using the lead (Pb) coolant in the primary system. NAVIS improves the economical efficiency resorting to the SCO 2 Brayton cycle for the secondary system. BORIS is operated by natural circulation of Pb without needing pumps. The reactor power is autonomously controlled by load-following operation without an active reactivity control system, whereas B 4 C based shutdown control rod is equipped for an emergency condition. SCO 2 promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the SCO 2 Brayton

  20. Magneto-hydrodynamically stable axisymmetric mirrorsa)

    Science.gov (United States)

    Ryutov, D. D.; Berk, H. L.; Cohen, B. I.; Molvik, A. W.; Simonen, T. C.

    2011-09-01

    Making axisymmetric mirrors magnetohydrodynamically (MHD) stable opens up exciting opportunities for using mirror devices as neutron sources, fusion-fission hybrids, and pure-fusion reactors. This is also of interest from a general physics standpoint (as it seemingly contradicts well-established criteria of curvature-driven instabilities). The axial symmetry allows for much simpler and more reliable designs of mirror-based fusion facilities than the well-known quadrupole mirror configurations. In this tutorial, after a summary of classical results, several techniques for achieving MHD stabilization of the axisymmetric mirrors are considered, in particular: (1) employing the favorable field-line curvature in the end tanks; (2) using the line-tying effect; (3) controlling the radial potential distribution; (4) imposing a divertor configuration on the solenoidal magnetic field; and (5) affecting the plasma dynamics by the ponderomotive force. Some illuminative theoretical approaches for understanding axisymmetric mirror stability are described. The applicability of the various stabilization techniques to axisymmetric mirrors as neutron sources, hybrids, and pure-fusion reactors are discussed; and the constraints on the plasma parameters are formulated.

  1. Magneto-hydrodynamically stable axisymmetric mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D. D.; Cohen, B. I.; Molvik, A. W. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Berk, H. L. [University of Texas, Austin, Texas 78712 (United States); Simonen, T. C. [University of California, Berkeley, California 94720 (United States)

    2011-09-15

    Making axisymmetric mirrors magnetohydrodynamically (MHD) stable opens up exciting opportunities for using mirror devices as neutron sources, fusion-fission hybrids, and pure-fusion reactors. This is also of interest from a general physics standpoint (as it seemingly contradicts well-established criteria of curvature-driven instabilities). The axial symmetry allows for much simpler and more reliable designs of mirror-based fusion facilities than the well-known quadrupole mirror configurations. In this tutorial, after a summary of classical results, several techniques for achieving MHD stabilization of the axisymmetric mirrors are considered, in particular: (1) employing the favorable field-line curvature in the end tanks; (2) using the line-tying effect; (3) controlling the radial potential distribution; (4) imposing a divertor configuration on the solenoidal magnetic field; and (5) affecting the plasma dynamics by the ponderomotive force. Some illuminative theoretical approaches for understanding axisymmetric mirror stability are described. The applicability of the various stabilization techniques to axisymmetric mirrors as neutron sources, hybrids, and pure-fusion reactors are discussed; and the constraints on the plasma parameters are formulated.

  2. Operational and passive safety aspects of the STAR-LM natural convection HLMC reactor. Study on operational aspects of a natural circulation HLMC reactor. 2

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Petkov, P.V.

    2001-09-01

    -in-turbine load and loss-of-heat sink without scram, the effects of uncertainties inherent in the calculation of the reactivity feedback coefficients are investigated. The calculations demonstrate that the reactor system exhibits the behavior required for stable startup and shutdown, autonomous load following, and passive safety. (author)

  3. System for unattended surveillance of nuclear reactor behavior

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.

    1977-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and dimensionality reduction capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns and to recognize deviations from these patterns were evaluated by experiments at the ORNL High-Flux Isotope Reactor. Power perturbations of less than 0.1% of the mean value in selected frequency ranges were readily detected by the pattern recognition system

  4. Consequences of the improvement of fast reactor material behavior under irradiation on fuel element performance

    International Nuclear Information System (INIS)

    Leclere, J.; Dupouy, J.M.; Marcon, J.P.

    1979-01-01

    The most important problems in fast reactor fuel element come from the excessive swelling of the structural materials used. The limitations of irradiation time for a given reactor result from the cladding or hexagonal wrapper deformations. Irradiation creep plays a major role, either in inducing additional deformations, or in providing possible ways of accommodation of bending stresses. Progress has been made in designing swelling resistant and/or low irradiation creep modulus materials. For instance in FRANCE, annealed 316 SS has been eliminated from pin and subassembly, and replaced by cold worked 316; we are now considering introduction of stabilizing elements in 316 SS as a further improvement and studying different alloys (nickel alloys, or ferritic steels). It has to be checked that the improvement of irradiation characteristic is not counterbalanced by losses on other properties (embrittlement for instance). Considering that pushing off or eliminating a limit may lead to the onset of a new one, it is porposed to make a review of the consequences of substantial improvement of structural material behavior

  5. 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

  6. Seismic response of the 'Cut-and Cover' type reactor containment considering nonlinear soil behavior

    International Nuclear Information System (INIS)

    El-Tahan, H.; Reddy, D.V.

    1979-01-01

    This paper describes some parametric studies of dynamic soil-structure interaction for the 'cut-and-cover' reactor concept. The dynamic loading considered is a horizontal earthquake motion. The high frequency ranges, which must be considered in the study of soil-structure interaction for nuclear power plants, and the nonlinearity of soil behavior during strong earthquakes are adequately taken into account. Soil nonlinearity is accounted for in an approximate manner using a combination of the 'equivalent linear method' and the method of complex response with complex moduli. The structure considered is a reinforced concrete containment for a 1100 - MWe power plant, buried in a dense sand medium. (orig.)

  7. Biofilm reactors for ethanol production

    Energy Technology Data Exchange (ETDEWEB)

    Vega, J L; Clausen, E C; Gaddy, J L

    1988-07-01

    Whole cell immobilization has been studied in the laboratory during the last few years as a method to improve the performance and economics of most fermentation processes. Among the various techniques available for cell immobilization, methods that provide generation of a biofilm offer reduced diffusional resistance, high productivities, and simple operation. This paper reviews some of the important aspects of biofilm reactors for ethanol production, including reactor start-up, steady state behavior, process stability, and mathematical modeling. Special emphasis is placed on covalently bonded Saccharomyces cerevisiae in packed bed reactors.

  8. Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitária, São Paulo (Brazil)

    2014-11-11

    This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model.

  9. Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Gonnelli, Eduardo; Diniz, Ricardo

    2014-01-01

    This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model

  10. 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

  11. State of the art of toshiba maintenance techniques for reactor internals

    International Nuclear Information System (INIS)

    Maekawa, Osamu; Hattori, Yasuhiro; Sudo, Akira

    2002-01-01

    As the number of aged plants increases, maintaining the integrity of the reactor pressure vessel and reactor internals in aged plants has become an essential issue to ensure continued stable operation and achieve higher plant operability. A major issue with regard to reactor internals is stress corrosion cracks (SCCs). Laser-applying techniques have many features suitable for preventive maintenance work on reactor internals. Toshiba has developed various laser-applying preventive maintenance techniques and accumulated considerable field experience utilizing these techniques in various aged plants. Moreover, in view of the importance of confirming the soundness of reactor internals in aged plants, Toshiba has developed and applied sophisticated nondestructive testing techniques for this purpose. (author)

  12. Isotopic alloying to tailor helium production rates in mixed spectrum reactors

    International Nuclear Information System (INIS)

    Mansur, L.K.; Rowcliffe, A.F.; Grossbeck, M.L.; Stoller, R.E.

    1985-01-01

    The purposes of this work are to increase the understanding of mechanisms by which helium affects microstructure and properties, to aid in the development of materials for fusion reactors, and to obtain data from fission reactors in regimes of direct interest for fusion reactor applications. Isotopic alloying is examined as a means of manipulating the ratio of helium transmutations to atom displacements in mixed spectrum reactors. The application explored is based on artificially altering the relative abundances of the stable isotopes of nickel to systematically vary the fraction of 58 Ni in nickel bearing alloys. The method of calculating helium production rates is described. Results of example calculations for proposed experiments in the High Flux Isotope Reactor are discussed

  13. Final Stage Development of Reactor Console Simulator

    International Nuclear Information System (INIS)

    Mohamad Idris Taib; Ridzuan Abdul Mutalib; Zareen Khan Abdul Jalil Khan; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Nurfarhana Ayuni Joha

    2013-01-01

    The Reactor Console Simulator PUSPATI TRIGA Reactor was developed since end of 2011 and now in the final stage of development. It is will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behavior and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of human system interface (HSI) is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate and estimated reactor console parameters. The capabilities in user interface, reactor physics and thermal-hydraulics can be expanded and explored to simulation as well as modeling for New Reactor Console, Research Reactor and Nuclear Power Plant. (author)

  14. RF-driven tokamak reactor with sub-ignited, thermally stable operation

    International Nuclear Information System (INIS)

    Harten, L.P.; Bers, A.; Fuchs, V.; Shoucri, M.M.

    1981-02-01

    A Radio-Frequency Driven Tokamak Reactor (RFDTR) can use RF-power, programmed by a delayed temperature measurement, to thermally stabilize a power equilibrium below ignition, and to drive a steady state current. We propose the parameters for such a device generating approx. = 1600 MW thermal power and operating with Q approx. = 40 (= power out/power in). A one temperature zero-dimensional model allows simple analytical formulation of the problem. The relevance of injected impurities for locating the equilibrium is discussed. We present the results of a one-dimensional (radial) code which includes the deposition of the supplementary power, and compare with our zero-dimensional model

  15. 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

  16. Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

    International Nuclear Information System (INIS)

    Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

    1999-02-01

    Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

  17. 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.)

  18. 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.

  19. Fusion reactor materials

    International Nuclear Information System (INIS)

    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

  20. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  1. Stable crack growth behaviors in welded CT specimens -- finite element analyses and simplified assessments

    International Nuclear Information System (INIS)

    Yagawa, Genki; Yoshimura, Shinobu; Aoki, Shigeru; Kikuchi, Masanori; Arai, Yoshio; Kashima, Koichi; Watanabe, Takayuki; Shimakawa, Takashi

    1993-01-01

    The paper describes stable crack growth behaviors in welded CT specimens made of nuclear pressure vessel A533B class 1 steel, in which initial cracks are placed to be normal to fusion line. At first, using the relations between the load-line displacement (δ) and the crack extension amount (Δa) measured in experiments, the generation phase finite element crack growth analyses are performed, calculating the applied load (P) and various kinds of J-integrals. Next, the simplified crack growth analyses based on the GE/EPRI method and the reference stress method are performed using the same experimental results. Some modification procedures of the two simplified assessment schemes are discussed to make them applicable to inhomogeneous materials. Finally, a neural network approach is proposed to optimize the above modification procedures. 20 refs., 13 figs., 1 tab

  2. 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)

  3. Simulation development for TRIGA reactor

    International Nuclear Information System (INIS)

    Handoyo, D.

    1997-01-01

    A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable

  4. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

    2013-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  6. The tempered one-sided stable density: a universal model for hydrological transport?

    International Nuclear Information System (INIS)

    Cvetkovic, Vladimir

    2011-01-01

    A generalized distribution for the water residence time in hydrological transport is proposed in the form of the tempered one-sided stable (TOSS) density. It is shown that limiting cases of the TOSS distribution recover virtually all distributions that have been considered in the literature for hydrological transport, from plug flow to flow reactor, the advection-dispersion model, and the gamma and Levy densities. The stable property of TOSS is particularly important, enabling a seamless transition between a time-domain random walk, and the Lagrangian (trajectory) approach along hydrological transport pathways.

  7. Strategies for the startup of methanogenic inverse fluidized-bed reactors using colonized particles.

    Science.gov (United States)

    Alvarado-Lassman, A; Sandoval-Ramos, A; Flores-Altamirano, M G; Vallejo-Cantú, N A; Méndez-Contreras, J M

    2010-05-01

    One of the inconveniences in the startup of methanogenic inverse fluidized-bed reactors (IFBRs) is the long period required for biofilm formation and stabilization of the system. Previous researchers have preferred to start up in batch mode to shorten stabilization times. Much less work has been done with continuous-mode startup for the IFBR configuration of reactors. In this study, we prepared two IFBRs with similar characteristics to compare startup times for batch- and continuous-operation modes. The reactors were inoculated with a small quantity of colonized particles and run for a period of 3 months, to establish the optimal startup strategy using synthetic media as a substrate (glucose as a source of carbon). After the startup stage, the continuous- and batch-mode reactors removed more than 80% of the chemical oxygen demand (COD) in 51 and 60 days of operation, respectively; however, at the end of the experiments, the continuous-mode reactor had more biomass attached to the support media than the batch-mode reactor. Both reactors developed fully covered support media, but only the continuous-mode reactor had methane yields close to the theoretical value that is typical of stable reactors. Then, a combined startup strategy was proposed, with industrial wastewater as the substrate, using a sequence of batch cycles followed by continuous operation, which allows stable operation at an organic loading rate of 20 g COD/L x d in 15 days. Using a fraction of colonized support as an inoculum presents advantages, with respect to previously reported strategies.

  8. Fracture mechanics analysis of reactor pressure vessel under pressurized thermal shock - The effect of elastic-plastic behavior and stainless steel cladding -

    International Nuclear Information System (INIS)

    Joo, Jae Hwang; Kang, Ki Ju; Jhung, Myung Jo

    2002-01-01

    Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). The PTS event means an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored

  9. 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.

  10. Simulation for Remote Operation for REX10 Nuclear Reactor

    International Nuclear Information System (INIS)

    Lee, Sim Won; Kim, Dong Su; Na, Man Gyun; Lee, Yoon Joon; Lee, Yeon Gun; Park, Goon Cherl

    2010-01-01

    The newly designed REX10 (Regional Energy Reactor, 10MWth) is an environmentally-friendly and stable small nuclear reactor for a small-scale reactor based Multi-purpose regional energy system. The REX10 has been developed to maintain system safety in order to be placed in densely populated region, island, etc. In addition, it is significantly hard to recruit many operation and maintenance personnel for small power reactors differently from usual commercial reactors because of its remote location and of economic reasons. In order to overcome these constraints, to decrease the operation cost by reducing operation and maintenance personnel, and to increase plant reliability through autonomous plant control, it is needed to design the control system of the small power reactors and to establish its unmanned remote operation system. In this study, the REX10 reactor core thermal power controller is designed by using a REX10 code analyzer. The remote control facility through man-machine interface (MMI) design and interface between programming languages was established and it was used to verify remote operation of REX10

  11. Cirus reactor: a milestone in Indian Atomic Energy Programme

    International Nuclear Information System (INIS)

    Ranjan, Rakesh; Karhadkar, C.G.; Bhattacharya, S.

    2017-01-01

    Cirus, a 40 MW_t_h, high flux, thermal neutron research reactor achieved first criticality on 10"t"h July 1960. It had vertical core, natural metallic Uranium rods in Aluminium clad as fuel, demineralised light water as coolant, heavy water as moderator, Helium as cover gas and graphite as reflector. A low-pressure containment was provided for the reactor and some of the important associated reactor systems. Reactor start-up and power regulation was effected by controlled adjustment of moderator level in the reactor vessel. Boron carbide rods were used as primary shut down devices. Dumping of heavy water from core worked as secondary shut down device. Seawater was used as secondary coolant for removal of the fission heat of the reactor. Initial operation of Cirus was marred by several difficulties, primarily arising out of water chemistry in primary cooling water system. It took almost 3 years to systematically resolve these problems and achieve stable operation of reactor. Cirus could be operated at its rated power by the year 1963

  12. 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

  13. 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.

  14. Predicted irradiation behavior of U3O8-Al dispersion fuels for production reactor applications

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Rest, J.

    1990-01-01

    Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U 3 O 8 -Al dispersion fuels. The U 3 O 8 -Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U 3 O 8 -Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U 3 O 8 -Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U 3 O 8 -Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U 3 O 8 -Al performance over a wide range of irradiation conditions

  15. Local bifurcation analysis in nuclear reactor dynamics by Sotomayor’s theorem

    International Nuclear Information System (INIS)

    Pirayesh, Behnam; Pazirandeh, Ali; Akbari, Monireh

    2016-01-01

    Highlights: • When the feedback reactivity is considered as a nonlinear function some complex behaviors may emerge in the system such as local bifurcation phenomenon. • The qualitative behaviors of a typical nuclear reactor near its equilibrium points have been studied analytically. • Comprehensive analytical bifurcation analyses presented in this paper are transcritical bifurcation, saddle- node bifurcation and pitchfork bifurcation. - Abstract: In this paper, a qualitative approach has been used to explore nuclear reactor behaviors with nonlinear feedback. First, a system of four dimensional ordinary differential equations governing the dynamics of a typical nuclear reactor is introduced. These four state variables are the relative power of the reactor, the relative concentration of delayed neutron precursors, the fuel temperature and the coolant temperature. Then, the qualitative behaviors of the dynamical system near its equilibria have been studied analytically by using local bifurcation theory and Sotomayor’s theorem. The results indicated that despite the uncertainty of the reactivity, we can analyze the qualitative behavior changes of the reactor from the bifurcation point of view. Notably, local bifurcations that were considered in this paper include transcritical bifurcation, saddle-node bifurcation and pitchfork bifurcation. The theoretical analysis showed that these three types of local bifurcations may occur in the four dimensional dynamical system. In addition, to confirm the analytical results the numerical simulations are given.

  16. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  17. Behavior of actinides in the Integral Fast Reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-06-01

    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 ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 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 plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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 ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 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.

  19. 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

  20. Behavior of actinides in the Integral Fast Reactor fuel cycle

    International Nuclear Information System (INIS)

    Courtney, J.C.; Lineberry, M.J.

    1994-01-01

    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 plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 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

  1. Expected characteristics of future reactors for human beings

    International Nuclear Information System (INIS)

    Taketani, Kiyoaki

    1992-01-01

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

  2. Decommissioning of multiple-reactor stations: facilitation by sequential decommissioning

    International Nuclear Information System (INIS)

    Moore, E.B.; Smith, R.I.; Wittenbrock, N.G.

    1982-01-01

    Reductions in cost and radiation dose can be achieved for decommissionings at multiple reactor stations because of factors not necessarily present at a single reactor station: reactors of similar design, the opportunity for sequential decommissioning, a site dedicated to nuclear power generation, and the option of either interim or permanent low-level radioactive waste storage facilities onsite. The cost and radiation dose reductions occur because comprehensive decommissioning planning need only be done once, because the labor force is stable and need only be trained once, because there is less handling of radioactive wastes, and because central stores, equipment, and facilities may be used. The cost and radiation dose reductions are sensitive to the number and types of reactors on the site, and to the alternatives selected for decommissioning. 3 tables

  3. TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

    International Nuclear Information System (INIS)

    Chalhoub, E.S.

    1980-09-01

    A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

  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. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  6. The reactor trainer: state-of-the-art classroom learning

    International Nuclear Information System (INIS)

    Stater, R.G.

    1996-01-01

    The Reactor Trainer is a professional, PC based, graphically enhanced, training resource specifically developed and customized for Class Room teaching of and learning about, reactor behavior. This unique, and focused, learning-target sets The Trainer apart from the panorama of the more common PC plant simulator. Its educational scope extends along a logical learning path, starting with important fundamental behavioral concepts of delayed neutrons, neutron multiplying factors, and reactor rate, moving to simple reactor transients in real time, and culminating with more complex operational evolutions. The Trainer empowers the Instructor with a dynamic Class Room demonstrator and the student with a superior hands-on learning tool. The Trainer's versatility encompasses a wide variety of educational needs, including initial operator training, requalification training, Shift Technical Advisor training, and other advanced or specialized training. In addition, The Reactor Trainer enhances prerequisite preparation of operator candidates for full-scale control room training and, in so doing, PC economics relieves full-scale simulator hours. (author)

  7. Research reactors for the social safety and prosperous neutron use

    International Nuclear Information System (INIS)

    Ito, Yasuo

    2000-01-01

    The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

  8. Xenon-induced power oscillations in a generic small modular reactor

    Science.gov (United States)

    Kitcher, Evans Damenortey

    As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a

  9. Effect of ship motions and flow stability in a small marine reactor driven by natural circulation

    International Nuclear Information System (INIS)

    Yoritsune, Tsutomu; Ishida, Toshihisa

    2001-12-01

    By using a small reactor as a power source for investigations and developments under sea, widely expanded activity is expectable. In this case, as for a nuclear reactor, small-size and lightweightness, and simplification of a system are needed with the safety. In JAERI, very small reactors for submersible research vessel (Deep-sea Reactor DRX and submersible Compact Reactor SCR) have been designed on the basis of needs investigation of sea research. Although the reactor is a PWR type, self-pressurization and natural circulation system are adopted in a primary system for small size and lightweightness. The fluid flow condition of the reactor core outlet is designed to be the two-phase with a low quality. Although the flow of a primary system is the two-phase flow with a low quality, the density wave oscillation may occur according to operating conditions. Moreover, since there are ship motions of heaving (the vertical direction acceleration) etc., when a submersible research vessel navigates on the sea surface, the circulation flow of the primary system is directly influenced by this external force. In order to maintain stable operations of the reactor, it is necessary to clarify effects of the flow stability characteristic of the primary coolant system and the external force. Until now, as for the flow stability of a nuclear reactor itself, many research reports have been published including the nuclear-coupled thermal oscillation of BWRs such as LaSalle-2, WNP-2 etc. As for the effect of external force, it is reported that the acceleration change based on a seismic wave affects the reactor core flow and the reactor power in a BWR. On the other hand, also in a PWR, since adoption of natural circulation cooling is considered for a generation 4 reactor, it is thought that the margin of the reactor core flow stability becomes an important parameter in the design. The reactor coolant flow mentioned in this report is the two-phase natural circulation flow coupled with

  10. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  11. 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)

  12. 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.,.

  13. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    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.,

  14. 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)

  15. 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.

  16. 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.

  17. 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.)

  18. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    2010-07-01

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  19. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

  20. 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

  1. Hydrolysis of cellulose in a cellulase-bead fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Karube, I; Tanaka, S; Shirai, T; Suzuki, S

    1977-08-01

    Cellulase was immobilized in a collagen fibril matrix, and no leakage of cellulase from the collagen fibril matrix was observed. The immobilized cellulase was more stable than the native cellulase. The substrate cellulose was hydrolyzed quantitatively with immobilized cellulase. The final reaction product was identified as glucose. Immobilized cellulase was used in a fluidized bed reactor where the pressure drop of the fluidized bed reactor was low and constant. Cellulose was hydrolyzed to glucose by the cellulase-bead fluidized bed reactor. The minimum flow velocity (U/sub mf/) was 0.5 cm/sec and the optimum flow velocity of the cellulose hydrolysis was 1 cm/sec.

  2. 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

  3. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2015-05-15

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  4. Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

    International Nuclear Information System (INIS)

    Hadek, J.; Grundmann, U.

    1989-09-01

    Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

  5. Aqueous homogeneous suspension reactor project. Report over the 4th quarter and the year 1974

    Energy Technology Data Exchange (ETDEWEB)

    1975-07-01

    The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described.

  6. Optimization and control of a continuous polymerization reactor

    Directory of Open Access Journals (Sweden)

    L. A. Alvarez

    2012-12-01

    Full Text Available This work studies the optimization and control of a styrene polymerization reactor. The proposed strategy deals with the case where, because of market conditions and equipment deterioration, the optimal operating point of the continuous reactor is modified significantly along the operation time and the control system has to search for this optimum point, besides keeping the reactor system stable at any possible point. The approach considered here consists of three layers: the Real Time Optimization (RTO, the Model Predictive Control (MPC and a Target Calculation (TC that coordinates the communication between the two other layers and guarantees the stability of the whole structure. The proposed algorithm is simulated with the phenomenological model of a styrene polymerization reactor, which has been widely used as a benchmark for process control. The complete optimization structure for the styrene process including disturbances rejection is developed. The simulation results show the robustness of the proposed strategy and the capability to deal with disturbances while the economic objective is optimized.

  7. Recent advances on membranes and membrane reactors for hydrogen production

    NARCIS (Netherlands)

    Gallucci, F.; Fernandez Gesalaga, E.; Corengia, P.; Sint Annaland, van M.

    2013-01-01

    Membranes and membrane reactors for pure hydrogen production are widely investigated not only because of the important application areas of hydrogen, but especially because mechanically and chemically stable membranes with high perm-selectivity towards hydrogen are available and are continuously

  8. Analysis on flow characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin

    1997-06-01

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

  9. Development of a nuclear reactor control system simulator using virtual instruments

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

    2011-01-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  10. Development of a nuclear reactor control system simulator using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  11. 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)

  12. Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

    Science.gov (United States)

    Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

    2017-05-01

    In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

  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. The operation characteristics of biohydrogen production in continuous stirred tank reactor with molasses

    Energy Technology Data Exchange (ETDEWEB)

    Hong, C.; Wei, H.; Jie-xuan, D.; Xin, Y.; Chuan-ping, Y. [Northeast Forestry Univ., Harbin (China). School of Forestry; Li, Y.F. [Northeast Forestry Univ., Harbin (China). School of Forestry; Shanghai Univ. Engineering, Shanghai (China). College of Chemistry and Chemical Engineering

    2010-07-01

    The anaerobic fermentation biohydrogen production in a continuous stirred tank reactor (CSTR) was investigated as a means for treating molasses wastewater. The research demonstrated that the reactor has the capacity of continuously producing hydrogen in an initial biomass (as volatile suspension solids) of 17.74 g/L, temperature of approximately 35 degrees Celsius, hydraulic retention time of 6 hours. The reactor could begin the ethanol-type fermentation in 12 days and realize stable hydrogen production. The study also showed that the CSTR reactor has a favourable stability even with an organic shock loading. The hydrogen yield and chemical oxygen demand (COD) increased, as did the hydrogen content.

  15. Contribution of stable isotopes to the study of pharmacokinetics of magnesium salts; Apport des isotopes stables a l'etude de la pharmacocinetique de sels de magnesium

    Energy Technology Data Exchange (ETDEWEB)

    Benech, H

    1999-05-28

    The use of stable isotopes as labels is becoming an attractive tool for the study of magnesium behavior in humans. It has been used two stable isotopes of magnesium, {sup 25}Mg and {sup 26}Mg, to measure the absolute bioavailability of a pharmaceutical form of magnesium. (N.C.)

  16. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  17. Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

    International Nuclear Information System (INIS)

    Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

    1980-01-01

    It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited

  18. Transmutation of stable isotopes and deactivation of radioactive waste in growing biological systems

    International Nuclear Information System (INIS)

    Vysotskii, Vladimir I.; Kornilova, Alla A.

    2013-01-01

    Highlights: ► The phenomena of isotope transmutation in growing microbiological cultures were investigated. ► Transmutation in microbiological associations is 20 times more effective than in pure cultures. ► Transmutation of radioactive nuclei to stable isotopes in such associations was investigated. ► The most accelerated rate of Cs 137 to stable Ba 138 isotope transmutation was 310 days. ► “Microbiological deactivation” may be used for deactivation of Chernobyl and Fukushima areas. - Abstract: The report presents the results of qualifying examinations of stable and radioactive isotopes transmutation processes in growing microbiological cultures. It is shown that transmutation of stable isotopes during the process of growth of microbiological cultures, at optimal conditions in microbiological associations, is 20 times more effective than the same transmutation process in the form of “one-line” (pure) microbiological cultures. In the work, the process of direct, controlled decontamination of highly active intermediate lifetime and long-lived reactor isotopes (reactor waste) through the process of growing microbiological associations has been studied. In the control experiment (flask with active water but without microbiological associations), the “usual” law of nuclear decay applies, and the life-time of Cs 137 isotope was about 30 years. The most rapidly increasing decay rate, which occurred with a lifetime τ * ≈ 310 days (involving an increase in rate, and decrease in lifetime by a factor of 35 times) was observed in the presence of Ca salt in closed flask with active water contained Cs 137 solution and optimal microbiological association

  19. Deposition of radionuclides and stable elements in Tokai-mura

    Energy Technology Data Exchange (ETDEWEB)

    Ueno, Takashi; Amano, Hikaru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    This report presents the data of deposition of radionuclides (Sep. 1993-March 2001) and stable elements (Sep. 1993-Oct. 1995) in Tokai-mura. To evaluate the migration of radionuclides and stable elements from the atmosphere to the ground surface, atmospheric deposition samples were collected from Sep. 1993 to March 2001 with three basins (distance to grand surface were 1.5 m, 4 m, 10 m) set up in the enclosure of JAERI in Tokai-mura, Ibaraki-ken, Japan. Monthly samples were evaporated to dryness to obtain residual samples and measured with a well type Ge detector for {sup 7}Be, {sup 40}K, {sup 137}Cs and {sup 210}Pb. According to the analysis of radioactivity, clear seasonal variations with spring peaks of deposition weight (dry) and deposition amounts of all objective radionuclides were found. Correlation analysis of deposition data also showed that these radionuclides can be divided into two groups. A part of dried sample was irradiated to reactor neutrons at JRR-4 for determination of stable element's deposition. (author)

  20. Transient fission-product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

    1997-12-01

    Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

  1. 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

  2. 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)

  3. 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.

  4. Simulation test of PIUS-type reactor with large scale experimental apparatus

    International Nuclear Information System (INIS)

    Tamaki, M.; Tsuji, Y.; Ito, T.; Tasaka, K.; Kukita, Yutaka

    1995-01-01

    A large scale experimental apparatus for simulating the PIUS-type reactor has been constructed keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were performed. Experimental results were compared with those obtained by the small scale apparatus in JAERI. We have already reported the effectiveness of the feedback control for the primary loop pump speed (PI control) for the stable operation. In this paper this feedback system is modified and the PID control is introduced. This new system worked well for the operation of the PIUS-type reactor even in a rapid transient condition. (author)

  5. A barrier on the public communication of nuclear technology. How to interpret reactor kinetics

    International Nuclear Information System (INIS)

    Yamamoto, Akio

    2007-01-01

    Reactor kinetics is very important to explain the safety of nuclear reactors. However, its description is somewhat complicated and not intuitive. In order to give more intuitive explanation for reactor kinetics, some metaphors that try to capture the feature of reactor behavior are discussed. (author)

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless under operating conditions of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Hong, Seok Min; Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Nuclear Materials Safety Research Division, Daejeon (Korea, Republic of); Kim, Seon Jin [Hanyang University, Division of materials science and engineering, Seoul (Korea, Republic of)

    2017-06-15

    The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

  11. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition; Cycle thorium et reacteurs a sel fondu: exploration du champ des parametres et des contraintes definissant le 'Thorium Molten Salt Reactor'

    Energy Technology Data Exchange (ETDEWEB)

    Mathieu, L

    2005-09-15

    Producing nuclear energy in order to reduce the anthropic CO{sub 2} emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  12. Measurements of low reactivities using a reactor oscillator

    International Nuclear Information System (INIS)

    Obradovic, D.; Petrovic, M.

    1965-12-01

    Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

  13. Measurements of low reactivities using a reactor oscillator

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D; Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-12-15

    Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

  14. Grey water treatment in UASB reactor at ambient temperature.

    Science.gov (United States)

    Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

    2007-01-01

    In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

  15. Contribution of stable isotopes to the study of pharmacokinetics of magnesium salts; Apport des isotopes stables a l'etude de la pharmacocinetique de sels de magnesium

    Energy Technology Data Exchange (ETDEWEB)

    Benech, H

    1999-05-28

    The use of stable isotopes as labels is becoming an attractive tool for the study of magnesium behavior in humans. It has been used two stable isotopes of magnesium, {sup 25}Mg and {sup 26}Mg, to measure the absolute bioavailability of a pharmaceutical form of magnesium. (N.C.)

  16. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  17. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    International Nuclear Information System (INIS)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

  18. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

    2012-01-01

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  19. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  20. Independent verification: operational phase liquid metal breeder reactors

    International Nuclear Information System (INIS)

    Bourne, P.B.

    1981-01-01

    The Fast Flux Test Facility (FFTF) recently achieved 100-percent power and now is in the initial stages of operation as a test reactor. An independent verification program has been established to assist in maintaining stable plant conditions, and to assure the safe operation of the reactor. Independent verification begins with the development of administrative procedures to control all other procedures and changes to the plant configurations. The technical content of the controlling procedures is subject to independent verification. The actual accomplishment of test procedures and operational maneuvers is witnessed by personnel not responsible for operating the plant. Off-normal events are analyzed, problem reports from other operating reactors are evaluated, and these results are used to improve on-line performance. Audits are used to confirm compliance with established practices and to identify areas where individual performance can be improved

  1. Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

  2. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Katagiri, Shigeo.

    1991-01-01

    The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

  3. Characteristics of a reactor with power reactivity feedback

    International Nuclear Information System (INIS)

    Li Fengyu; Zhang Yusheng; Zhang Guangfu; Liu Ying

    2008-01-01

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

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kebadze, B V; Adamovski, L A

    1975-06-01

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

  6. 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.)

  7. Real-time stability monitoring method for boiling water reactor nuclear power plants

    International Nuclear Information System (INIS)

    Fukunishi, K.; Suzuki, S.

    1987-01-01

    A method for real-time stability monitoring is developed for supervising the steady-state operation of a boiling water reactor core. The decay ratio of the reactor power fluctuation is determined by measuring only the output neutron noise. The concept of an inverse system is introduced to identify the dynamic characteristics of the reactor core. The adoption of an adaptive digital filter is useful in real-time identification. A feasibility test that used measured output noise as an indication of reactor power suggests that this method is useful in a real-time stability monitoring system. Using this method, the tedious and difficult work for modeling reactor core dynamics can be reduced. The method employs a simple algorithm that eliminates the need for stochastic computation, thus making the method suitable for real-time computation with a simple microprocessor. In addition, there is no need to disturb the reactor core during operation. Real-time stability monitoring using the proposed algorithm may allow operation under less stable margins

  8. 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)

  9. 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

  10. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  11. On the study of catalytic membrane reactor for water detritiation: Modeling approach

    Energy Technology Data Exchange (ETDEWEB)

    Liger, Karine, E-mail: karine.liger@cea.fr [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Mascarade, Jérémy [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Joulia, Xavier; Meyer, Xuan-Mi [Université de Toulouse, INPT, UPS, Laboratoire de Génie Chimique, 4, Allée Emile Monso, Toulouse F-31030 (France); CNRS, Laboratoire de Génie Chimique, Toulouse F-31030 (France); Troulay, Michèle; Perrais, Christophe [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France)

    2016-11-01

    Highlights: • Experimental results for the conversion of tritiated water (using deuterium as a simulant of tritium) by means of a catalytic membrane reactor in view of tritium recovery. • Phenomenological 2D model to represent catalytic membrane reactor behavior including the determination of the compositions of gaseous effluents. • Good agreement between the simulation results and experimental measurements performed on the dedicated facility. • Explanation of the unexpected behavior of the catalytic membrane reactor by the modeling results and in particular the gas composition estimation. - Abstract: In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium bonded water can be recovered under the valuable Q{sub 2} form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal of reactions products all along the reactor, and thus limits reverse reaction rate to the benefit of the direct one (shift effect). The reactions kinetics, which are still little known or unknown, are generally assumed to be largely greater than the permeation ones so that thermodynamic equilibriums of isotope exchange reactions are generally assumed. This paper proposes a new phenomenological 2D model to represent catalytic membrane reactor behavior with the determination of gas effluents compositions. A good agreement was obtained between the simulation results and experimental measurements performed on a dedicated facility. Furthermore, the gas composition estimation permits to interpret unexpected behavior of the catalytic membrane reactor. In the next future, further sensitivity analysis will be performed to determine the limits of the model and a kinetics study will be conducted to assess the thermodynamic equilibrium of reactions.

  12. On the study of catalytic membrane reactor for water detritiation: Modeling approach

    International Nuclear Information System (INIS)

    Liger, Karine; Mascarade, Jérémy; Joulia, Xavier; Meyer, Xuan-Mi; Troulay, Michèle; Perrais, Christophe

    2016-01-01

    Highlights: • Experimental results for the conversion of tritiated water (using deuterium as a simulant of tritium) by means of a catalytic membrane reactor in view of tritium recovery. • Phenomenological 2D model to represent catalytic membrane reactor behavior including the determination of the compositions of gaseous effluents. • Good agreement between the simulation results and experimental measurements performed on the dedicated facility. • Explanation of the unexpected behavior of the catalytic membrane reactor by the modeling results and in particular the gas composition estimation. - Abstract: In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium bonded water can be recovered under the valuable Q_2 form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal of reactions products all along the reactor, and thus limits reverse reaction rate to the benefit of the direct one (shift effect). The reactions kinetics, which are still little known or unknown, are generally assumed to be largely greater than the permeation ones so that thermodynamic equilibriums of isotope exchange reactions are generally assumed. This paper proposes a new phenomenological 2D model to represent catalytic membrane reactor behavior with the determination of gas effluents compositions. A good agreement was obtained between the simulation results and experimental measurements performed on a dedicated facility. Furthermore, the gas composition estimation permits to interpret unexpected behavior of the catalytic membrane reactor. In the next future, further sensitivity analysis will be performed to determine the limits of the model and a kinetics study will be conducted to assess the thermodynamic equilibrium of reactions.

  13. Decreased activity of a propionate degrading community in a UASB reactor fed with synthetic medium without molybdenum, tungsten and selenium

    NARCIS (Netherlands)

    Worm, P.; Fermoso, F.G.; Lens, P.N.L.; Plugge, C.M.

    2009-01-01

    The composition and dynamics of the propionate degrading community in a propionate-fed upflow anaerobic sludge bed (UASB) reactor with sludge originating from an alcohol distillery wastewater treating UASB reactor was studied. The rather stable propionate degrading microbial community comprised

  14. Self-teaching neural network learns difficult reactor control problem

    International Nuclear Information System (INIS)

    Jouse, W.C.

    1989-01-01

    A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits

  15. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  16. 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

  17. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

    International Nuclear Information System (INIS)

    Was, Gary S.

    2007-01-01

    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

  18. 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.)

  19. Contribution of stable isotopes to the study of pharmacokinetics of magnesium salts

    International Nuclear Information System (INIS)

    Benech, H.

    1999-01-01

    The use of stable isotopes as labels is becoming an attractive tool for the study of magnesium behavior in humans. It has been used two stable isotopes of magnesium, 25 Mg and 26 Mg, to measure the absolute bioavailability of a pharmaceutical form of magnesium. (N.C.)

  20. Is Sensation Seeking a Stable Trait or Does It Change over Time?

    Science.gov (United States)

    Lynne-Landsman, Sarah D.; Graber, Julia A.; Nichols, Tracy R.; Botvin, Gilbert J.

    2011-01-01

    The theory of sensation seeking has conceptualized this construct as a stable personality trait associated with a variety of problem behaviors. Reckless behavior theory posits that increases in reckless behavior during adolescence can be attributed, in part, to increases in sensation seeking. This study evaluated patterns of stability and change…

  1. Two-year microbial adaptation during hydrogen-mediated biogas upgrading process in a serial reactor configuration

    DEFF Research Database (Denmark)

    Treu, Laura; Kougias, Panagiotis; de Diego-Díaz, B.

    2018-01-01

    Microbial dynamics in an upgrading biogas reactor system undergoing a more than two years-period at stable operating conditions were explored. The carbon dioxide generated during biomass degradation in the first reactor of the system was converted to methane into the secondary reactor by addition...... of external hydrogen. Considering the overall efficiency, the long-term operation period resulted in an improved biogas upgrading performance (99% methane content). However, a remarkable accumulation of acetate was revealed, indicating the enhancement of homoacetogenic activity. For this reason, a shift...

  2. The cognitive-behavioral system of leadership: cognitive antecedents of active and passive leadership behaviors

    Science.gov (United States)

    Dóci, Edina; Stouten, Jeroen; Hofmans, Joeri

    2015-01-01

    In the present paper, we propose a cognitive-behavioral understanding of active and passive leadership. Building on core evaluations theory, we offer a model that explains the emergence of leaders’ active and passive behaviors, thereby predicting stable, inter-individual, as well as variable, intra-individual differences in both types of leadership behavior. We explain leaders’ stable behavioral tendencies by their fundamental beliefs about themselves, others, and the world (core evaluations), while their variable, momentary behaviors are explained by the leaders’ momentary appraisals of themselves, others, and the world (specific evaluations). By introducing interactions between the situation the leader enters, the leader’s beliefs, appraisals, and behavior, we propose a comprehensive system of cognitive mechanisms that underlie active and passive leadership behavior. PMID:26441721

  3. Premature and stable critical heat flux for downward flow in a narrow rectangular channel

    International Nuclear Information System (INIS)

    Lee, Juhyung; Chang, Soon Heung; Jeong, Yong Hoon; Jo, Daeseong

    2014-01-01

    It has been recommended that RRs and MTRs be designed to have sufficient margins for CHF and the onset of FI as well, since unstable flow could leads to premature CHF under very low wall heat flux in comparison to stable CHF. Even the fact and previous studies, however, the understanding of relationship among FI, premature CHF and stable CHF is not sufficient to date. In this regards, subcooled flow boiling in a vertical rectangular channel was experimentally investigated to enhance the understanding of the CHF and the effect of the two-phase flow instability on it under low pressure conditions, especially for downward flow which was adopted for Jordan Research and Training Reactor (JRTR) and Kijang research reactor (KJRR) to achieve easier fuel and irradiation rig loading. In this study, CHF for downward flow of water under low pressure in narrow rectangular channel was experimentally investigated. For conditions such as downward flow, narrow rectangular channel and low pressure, it has been deduced from literature that flow instability could largely influence on triggering CHF at lower heat flux, i. e. premature CHF. Total 54 CHF data, which includes premature and stable data was obtained for various fluid conditions and system configurations including inlet stiffness. The upper and lower boundaries of CHF were newly proposed based on the experiment

  4. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  5. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  6. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  7. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    Science.gov (United States)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

    The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni-Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

  8. Time to reach a given level of number of neutrons is stochastic analog of reactor period

    International Nuclear Information System (INIS)

    Ryazanov, V.V.

    2012-01-01

    In theory and in practice the operation of nuclear reactors to control the safety of the reactor is widely used deterministic value - the period of the reactor. It is proposed along with the period of the reactor using a stochastic analogue of this magnitude - a random amount of time to achieve a given level of a random process for the number of neutrons in the reactor. The paper discusses various features of the behavior of the mean and variance of time to achieve a specified level. This kind of features can be associated with impaired behavior of the reactor system. Introduced the value of time required to reach the level can be used to monitor and improve the safety of nuclear power plants

  9. Impact of confinement physics on reactor design and economics

    International Nuclear Information System (INIS)

    DeFreece, D.A.; Campbell, R.B.; Waganer, L.M.

    1977-01-01

    A variety of confinement laws were employed in a transient, zero dimensional plasma code, which was coupled to the TOCOMO systems code. The purpose was to determine the impact of the confinement laws on reactor design, power costs and changes in the utility interface. A satisfactory reactor and power plant has been defined for the large majority of combinations of confinement law, power plant size and plasma shape. Trapped ion mode (TIM) has been the easiest to work with, since the plasma is thermally stable with a good power density and minimal alpha particle build up. Neoclassical and pseudoclassical along with TEMII result in satisfactory reactor performance, but require active feedback control (by injecting impurities) to prevent plasma temperature excursions. These laws also require some form and degree of confinement time spoiling to allow long burn times, otherwise, alpha particles build up to an unacceptable level. TEM I results in thermal equilibrium at 5 keV and must be driven to provide a reactor quality plasma. The continuous injected power required for a 4300 MW thermal reactor is 540 MW. This added to the other circulating loads results in a net power output of 600 MWe at a very high relative cost. Daughney (empirical) confinement results in a satisfactory, competitive reactor

  10. Utilization of fast reactor excess neutrons for burning long-lived fission products

    International Nuclear Information System (INIS)

    Kawashima, K.; Kobayashi, K.; Kaneto, K.

    1995-01-01

    An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

  11. Thermophilic two-phase anaerobic digestion using an innovative fixed-bed reactor for enhanced organic matter removal and bioenergy recovery from sugarcane vinasse

    International Nuclear Information System (INIS)

    Fuess, Lucas Tadeu; Kiyuna, Luma Sayuri Mazine; Ferraz, Antônio Djalma Nunes; Persinoti, Gabriela Felix

    2017-01-01

    Highlights: • An innovative fixed-film anaerobic reactor was applied to sugarcane vinasse. • Stable operation was observed for OLRs as high as 30 kg COD m"−"3 day"−"1. • Propionate buildup did not impact the stability of the structured-bed reactor. • Enhanced bioenergy recovery was estimated from biodigestion with phase separation. • Energy extraction was over 20% higher compared to single-phase systems. - Abstract: This study considered the application of anaerobic digestion (AD) with phase separation combined with the use of an anaerobic structured-bed reactor (ASTBR) as the methanogenic phase for the treatment of sugarcane vinasse, a high-strength wastewater resulting from ethanol production. Two combined thermophilic acidogenic-methanogenic systems formed by one single acidogenic reactor followed by two methanogenic reactors operated in parallel were compared, namely, a conventional UASB reactor and an upflow ASTBR reactor. Increasing organic loading rate (OLR) conditions (15–30 kg COD m"−"3 d"−"1) were applied to the methanogenic reactors. The results highlighted the feasibility of applying the ASTBR to vinasse, indicating a global COD removal higher than 80%. The ASTBR exhibited a stable long-term operation (240 days), even for OLR values as high as 30 kg COD m"−"3 d"−"1. The application of similar conditions to the UASB reactor indicated severe performance losses, leading to the accumulation of acids for every increase in the OLR. An energetic potential of 181.5 MJ for each cubic meter of vinasse was estimated from both hydrogen and methane. The provision of bicarbonate alkalinity proved to be a key factor in obtaining stable performance, offsetting the limitations of relatively low hydraulic retention times (<24 h).

  12. The utility industry and reactor surveillance

    International Nuclear Information System (INIS)

    Jenkins, R.B.

    1983-01-01

    Every commercial nuclear power reactor pressure vessel (RPV) is required to have a reactor vessel surveillance program at the time of plant licensing. The program is part of a continuing structural integrity assessment of the RPV. As such, the surveillance program supplements Section III of the American Society of Mechanical Engineers (ASME) Code (1), which is the design basis for nuclear power plant component pressure boundaries. The Code assumes that the materials of construction are ductile in the evaluation and design of all components. The surveillance program for each RPV is intended to provide assurance of continued applicability of the ASME Code, Appendix G, assessment of that RPV's operating limits. This assessment ensures that the RPV is always in a condition which precludes the unstable propagation of flaws in the vessel wall material. The potential presence of flaws and the desire to ensure ductility are significant considerations in ferritic steels such as those used to fabricate nuclear reactor pressure vessels. These materials are known to exhibit transition from ductile-to-brittle fracture behavior over a determined temperature range. Neutron irradiation tends to shift this ductile-to-brittle behavior transition zone to a temperature higher than unirradiated materials

  13. Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

    International Nuclear Information System (INIS)

    Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

    2013-01-01

    The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

  14. 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.

  15. Identification of Chemical Reactor Plant’s Mathematical Model

    OpenAIRE

    Pyakullya, Boris Ivanovich; Kladiev, Sergey Nikolaevich

    2015-01-01

    This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  16. Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1999-01-01

    The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

  17. MELCOR development for existing and advanced reactors

    International Nuclear Information System (INIS)

    Summers, R.M.

    1993-01-01

    Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensers, debris quenching, lower plenum debris behavior, core materials interactions' and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and fine-scale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR's capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and creep rupture failure of the lower head. Additional development needs in other areas are discussed

  18. RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

    This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

  19. Software for the nuclear reactor dynamics study using time series processing

    International Nuclear Information System (INIS)

    Valero, Esbel T.; Montesino, Maria E.

    1997-01-01

    The parametric monitoring in Nuclear Power Plant (NPP) permits the operational surveillance of nuclear reactor. The methods employed in order to process this information such as FFT, autoregressive models and other, have some limitations when those regimens in which appear strongly non-linear behaviors are analyzed. In last years the chaos theory has offered new ways in order to explain complex dynamic behaviors. This paper describes a software (ECASET) that allow, by time series processing from NPP's acquisition system, to characterize the nuclear reactor dynamic as a complex dynamical system. Here we show using ECASET's results the possibility of classifying the different regimens appearing in nuclear reactors. The results of several temporal series processing from real systems are introduced. This type of analysis complements the results obtained with traditional methods and can constitute a new tool for monitoring nuclear reactors. (author). 13 refs., 3 figs

  20. Modeling the electrochemistry of the primary circuits of light water reactors

    International Nuclear Information System (INIS)

    Bertuch, A.; Macdonald, D.D.; Pang, J.; Kriksunov, L.; Arioka, K.

    1994-01-01

    To model the corrosion behaviors of the heat transport circuits of light water reactors, a mixed potential model (NTM) has been developed and applied to both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Using the data generated by the GE/UKEA-Harwell radiolysis model, electrochemical potentials (ECPs) have been calculated for the heat transport circuits of eight BWRs operating under hydrogen water chemistry (HWC). By modeling the corrosion behaviors of these reactors, the effectiveness of HWC at limiting IGSCC and IASCC can be determined. For simulating PWR primary circuits, a chemical-radiolysis model (developed by the authors) was used to generate input parameters for the MPM. Corrosion potentials of Type 304 and 316 SSs in PWR primary environments were calculated using the NTM and were found to be in good agreement with the corrosion potentials measured in the laboratory for simulated PWR primary environments

  1. 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.

  2. Technology which led to the westinghouse inherently safe liquid metal reactor

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

    1985-01-01

    The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

  3. Warm Parenting Associated with Decreasing or Stable Child BMI during Treatment.

    Science.gov (United States)

    Rhee, Kyung E; Jelalian, Elissa; Boutelle, Kerri; Dickstein, Susan; Seifer, Ronald; Wing, Rena

    2016-04-01

    While authoritative parenting, which includes high levels of warmth and behavioral control, has been associated with lower risk of obesity, little is known about how general parenting impacts child weight loss during treatment. Our goal was to examine the relationship between several general parenting dimensions and 'decreasing /stable' child BMI during a 16-week family-based behavioral weight control program. Forty-four overweight parent-child dyads (child age 8 to 12 years) enrolled in the program. Families were videotaped at baseline eating dinner in their home. Using the General Parenting Observational Scale (GPOS), meals were coded for several general parenting dimensions. Primary outcome was percent of children whose BMI 'decreased or stayed the same.' Multivariable logistic regression was used to determine the relationship between general parenting and decreasing/stable child BMI. Forty families (91%) completed the program. Children had a mean BMI change of -0.40 (SD 1.57), which corresponds to a -0.15 (SD 0.20) change in BMI z-score (BMI-Z); 75% of children had decreasing/stable BMI. In the unadjusted models, lower parent BMI, higher parent education, and higher levels of parental warmth were significantly associated with decreasing/stable child BMI. In the multivariable model, only higher level of warmth was associated with increased odds of decreasing/stable child BMI (OR = 1.28; 95% CI, 1.01, 1.62). Baseline parental warmth may influence a child's ability to lower/maintain BMI during a standard family-based behavioral weight control program. Efforts to increase parent displays of warmth and emotional support towards their overweight child may help to increase the likelihood of treatment success.

  4. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  5. Software for the nuclear reactor dynamics study using time series processing; Software para el estudio de la dinamica de reactores nucleares mediante el procesamiento de series temporales

    Energy Technology Data Exchange (ETDEWEB)

    Valero, Esbel T.; Montesino, Maria E. [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

    1997-12-01

    The parametric monitoring in Nuclear Power Plant (NPP) permits the operational surveillance of nuclear reactor. The methods employed in order to process this information such as FFT, autoregressive models and other, have some limitations when those regimens in which appear strongly non-linear behaviors are analyzed. In last years the chaos theory has offered new ways in order to explain complex dynamic behaviors. This paper describes a software (ECASET) that allow, by time series processing from NPP`s acquisition system, to characterize the nuclear reactor dynamic as a complex dynamical system. Here we show using ECASET`s results the possibility of classifying the different regimens appearing in nuclear reactors. The results of several temporal series processing from real systems are introduced. This type of analysis complements the results obtained with traditional methods and can constitute a new tool for monitoring nuclear reactors. (author). 13 refs., 3 figs.

  6. 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)

  7. 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.

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

    International Nuclear Information System (INIS)

    Phillpot, Simon; Tulenko, James

    2011-01-01

    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.

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

    KAUST Repository

    Tao, Tao; Sun, Wenyu; Hansen, Nils; Jasper, Ahren W.; Moshammer, Kai; Chen, Bingjie; Wang, Zhandong; Huang, Can; Dagaut, Philippe; Yang, Bin

    2018-01-01

    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.

  11. 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.

  12. Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

    International Nuclear Information System (INIS)

    Schweizer, Fernando Lage Araújo

    2014-01-01

    The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

  13. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

  14. Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

    Science.gov (United States)

    Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

    2017-12-01

    The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

  15. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  16. Cobalt-60 production in CANDU reactors

    International Nuclear Information System (INIS)

    Ross, Michel; Lemire, Christian

    2002-01-01

    CANDU reactors can produce cobalt-60 very efficiently and with an interesting return on investment. This paper discusses what is needed to convert a CANDU reactor into a cobalt-60 producer: what are the different phases, the safety studies required, the physical modifications needed, and what is the minimum involvement of the utility owning the plant. The past ten years of experience of Hydro-Quebec as a cobalt-60 producer will be reviewed, including the management of the risk of both incident and electricity generation loss, and including the benefits for the utility and its personnel. Originally a simple metal used for centuries as a pigment, cobalt-59 today is transformed into cobalt-60, a radioactive element of unprecedented value. Well known in medicine for cancer treatment, cobalt-60 is also used to sterilize a wide range of disposable medical products used in hospitals and to sanitize pharmaceutical and cosmetic products. Cobalt-60 is proving to be a new and effective solution, in the food sector, for preserving harvests and controlling food-borne diseases, or to advantageously replace certain gases and chemical products which are suspected of being harmful or carcinogenic. There are also other applications, such as: hardening of some plastics, treatment of sewage sludge and elimination of harmful insect populations. With a half-life of 5,3 years, cobalt-60 is a metal not found in nature. It is a radioactive isotope produced by exposing stable nuclei of cobalt-59 to neutrons. One of the best places to find such an important neutron source is a nuclear reactor. High energy gamma rays are then emitted during the process of radioactive decay, where cobalt-60 seeks again its stable state

  17. IDAS-RR: an incident data base system for research reactors

    International Nuclear Information System (INIS)

    Matsumoto, Kiyoshi; Kohsaka, Atsuo; Kaminaga, Masanori; Murayama, Youji; Ohnishi, Nobuaki; Maniwa, Masaki.

    1990-03-01

    An Incident Data Base System for Research Reactors, IDAS-RR, has been developed. IDAS-RR has information about abnormal incidents (failures, transients, accidents, etc.) of research reactors in the world. Data reference, input, editing and other functions of IDAS-RR are menu driven. The routine processing and data base management functions are performed by the system software and hardware. PC-9801 equipment was selected as the hardware because of its portability and popularity. IDAS-RR provides effective reference information for the following activities. 1) Analysis of abnormal incident of research reactors, 2) Detail analysis of research reactor behavior in the abnormal incident for building the knowledge base of the reactor emergency diagnostic system for research reactor, 3) Planning counter-measure for emergency situation in the research reactor. This report is a user's manual of IDAS-RR. (author)

  18. Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Waloch, M.L.

    1977-01-01

    In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

  19. Chaotic and stable perturbed maps: 2-cycles and spatial models

    Science.gov (United States)

    Braverman, E.; Haroutunian, J.

    2010-06-01

    As the growth rate parameter increases in the Ricker, logistic and some other maps, the models exhibit an irreversible period doubling route to chaos. If a constant positive perturbation is introduced, then the Ricker model (but not the classical logistic map) experiences period doubling reversals; the break of chaos finally gives birth to a stable two-cycle. We outline the maps which demonstrate a similar behavior and also study relevant discrete spatial models where the value in each cell at the next step is defined only by the values at the cell and its nearest neighbors. The stable 2-cycle in a scalar map does not necessarily imply 2-cyclic-type behavior in each cell for the spatial generalization of the map.

  20. Tritium problems in fusion reactor systems

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1975-01-01

    A brief introduction is given to the role tritium will play in the development of fusion power. The biological and worldwide environmental behavior of tritium is reviewed. The tritium problems expected in fusion power reactors are outlined. A few thoughts on tritium permeation and recent results for tritium cleanup and CT 4 accumulation are presented. Problems involving the recovery of tritium from the breeding blanket in fusion power reactors are also considered, including the possible effect of impurities in lithium blankets and the use of lithium as a regenerable getter pump. (auth)

  1. 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

  2. Identification of Chemical Reactor Plant’s Mathematical Model

    Directory of Open Access Journals (Sweden)

    Pyakillya Boris

    2015-01-01

    Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  3. Thermochemical data for reactor materials and fission products

    International Nuclear Information System (INIS)

    Cordfunke, E.H.P.; Konings, R.J.M.

    1990-01-01

    This volume presents a collection of critically assessed data on inorganic compounds which are of special interest in nuclear reactor safety studies. Thermodynamic equilibrium calculations are an important and widely used instrument in the understanding of the chemical behavior and release of fission products in the course of nuclear reactor accidents. The reliability of such calculations is, nevertheless, limited by the availability of accurate input data for relevant compounds

  4. Ultras-stable Physical Vapor Deposited Amorphous Teflon Films with Extreme Fictive Temperature Reduction

    Science.gov (United States)

    McKenna, Gregory; Yoon, Heedong; Koh, Yung; Simon, Sindee

    In the present work, we have produced highly stable amorphous fluoropolymer (Teflon AF® 1600) films to study the calorimetric and relaxation behavior in the deep in the glassy regime. Physical vapor deposition (PVD) was used to produce 110 to 700 nm PVD films with substrate temperature ranging from 0.70 Tg to 0.90 Tg. Fictive temperature (Tf) was measured using Flash DSC with 600 K/s heating and cooling rates. Consistent with prior observations for small molecular weight glasses, large enthalpy overshoots were observed in the stable amorphous Teflon films. The Tf reduction for the stable Teflon films deposited in the vicinity of 0.85 Tg was approximately 70 K compared to the Tgof the rejuvenated system. The relaxation behavior of stable Teflon films was measured using the TTU bubble inflation technique and following Struik's protocol in the temperature range from Tf to Tg. The results show that the relaxation time decreases with increasing aging time implying that devitrification is occurring in this regime.

  5. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  6. 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

  7. Behavior of exposed human lymphocytes to a neutron beam of the Reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Carbajal R, M. I.; Arceo M, C.; Aguilar H, F.; Guerrero C, C.

    2012-10-01

    The living beings are permanently exposed to radiations of natural origin: cosmic and geologic, as well as the artificial radiations that come from sources elaborated by the man. The artificial sources have an important use in the medical area. Particularly has been increased the neutrons use due to the effectiveness that they have to damage the cells with regard to other radiation types. The biological indicator of exposition to ionizing radiation more reliable is the chromosomal aberrations study, specifically the dicentrics in human lymphocytes. This test allows, establishing the exposition dose in function of the damage quantity. The dicentrics have a behavior in function of the dose. The calibration curve that describes this behavior is specific for each type of ionizing radiation. In the year 2006 beginning was given to the expositions of human lymphocytes to a neutron beam generated in the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico. Up to 2008 the response dose curve comprised an interval of exposition time of up to 30 minutes. Moreover, the interval between 10 an 20 minutes is included, since was observed that this last is indispensable for the adjustment waited in a lineal model. (Author)

  8. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  9. Numerical model simulation of free surface behavior in spallation target of ADS

    International Nuclear Information System (INIS)

    Chai Xiang; Su Guanyu; Cheng Xu

    2012-01-01

    The spallation target in accelerator driven sub-critical system (ADS) couples the subcritical reactor core with accelerator. The design of a windowless target has to ensure the formation of a stable free surface with desirable shape, to avoid local over- heating of the heavy liquid metal (HLM). To investigate the free surface behavior of the spallation target, OpenFOAM, an opened CFD software platform, was used to simulate the formation and features of the free surface in the windowless target. VOF method was utilized as the interface-capturing methodology. The numerical results were compared to experimental data and numerical results obtained with FLUENT code. The effects of turbulence models were studied and recommendations were made related to application of turbulence models. (authors)

  10. Effect of core burnup on the dynamic behavior of fast reactors

    International Nuclear Information System (INIS)

    Ilberg, D.; Saphier, D.; Yiftah, S.

    1977-01-01

    Performance of a dynamic analysis, taking burnup changes into account, requires fission-product nuclear data of relatively small uncertainty, suitable burnup calculation models, and dynamic computer programs. These were prepared and used with the following results: (1) Significant changes in static and dynamic parameters were observed when investigating the effect of burnup. These changes were found to be larger than differences introduced by the uncertainty of the fission-product nuclear data. (2) A one-dimensional burnup computer program was prepared. It was found that a burnup model based on the generalized radioactive decay scheme is suitable for accurate fast reactor calculations. (3) Space-time dynamic calculations of fast reactors having different burnup levels were performed. The stability difference between ''clean'' and high burnup cores is greater when local rather than uniform perturbations are inserted along the entire core length. The magnitude by which the ''end-of-life'' core increases the transient excursion over that of the clean core depends on the particular region in which the perturbation is inserted. The end-of-life core will magnify the transient excursion more than the clean core whenever the perturbation is inserted into a region having a higher adjoint flux level than that of the clean core. However, when a reactor safety system operates successfully, the difference in the temperature transient of the clean and end-of-life cores will be relatively small. It is suggested that only the analysis of large local perturbations be performed for end-of-life cores as well as for clean cores in the safety evaluation of fast reactors

  11. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Wagner de Souza; Py Junior, Delcy de Azevedo; Goncalves, Simone, E-mail: wspereira@inb.gov.br [Unidade de Tratamento de Minerio (UTM/INB), Pocos de Caldas, MG (Brazil). Coordenacao de Protecao Radiologica. Grupo Multidisciplinar de Radioprotecao; Kelecom, Alphonse [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Inst. de Biologia. Lab. de Radiobiologia e Radiometria Pedro Lopes dos Santos; Morais, Gustavo Ferrari de; Campelo, Emanuele Lazzaretti Cordova [Unidade de Tratamento de Minerio (UTM/INB), Pocos de Caldas, MG (Brazil). Coordenacao de Desenvolvimento de Processos; Dores, Luis Augusto de Carvalho Bresser [Unidade de Tratamento de Minerio (UTM/INB), Pocos de Caldas, MG (Brazil). Gerencia de Descomissionamento

    2011-07-01

    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)

  13. Severe accidents and terrorist threats at nuclear reactors

    International Nuclear Information System (INIS)

    Pollack, G.L.

    1987-01-01

    Some of the key areas of uncertainty are the nature of the physical and chemical interactions of released fission products and of the interactions between a molten core and concrete, the completeness and validity of the computer codes used to predict accidents, and the behavior of the containment. Because of these and other uncertainties, it is not yet possible to reliably predict the consequences of reactor accidents. It is known that for many accident scenarios, especially less severe ones or where the containment is not seriously compromised, the amount of radioactive material expected to escape the reactor is less, even much less, than was previously calculated. For such accidents, the predictions are easier and more reliable. With severe accidents, however, there is considerable uncertainty as to the predicted results. For accidents of the type that terrorists might cause - for example, where the sequence of failure would be unexpected or where redundant safety features are caused to fail together - the uncertainties are still larger. The conclusion, then, is that there are potential dangers to the public from terrorist actions at a nuclear reactor; however, because of the variety of potential terrorist threats and the incompleteness of the knowledge about the behavior of reactor components and fission products during accidents, the consequences cannot yet be assessed quantitatively

  14. Development of an educational nuclear research reactor simulator

    International Nuclear Information System (INIS)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim; Ashoub, Nagieb

    2014-01-01

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  15. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  16. 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

  17. Study of power peak migration due to insertion of control bars in a PWR reactor

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Costa, Danilo Leite; Borges, Diogo da Silva; Lava, Deise Diana; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown

  18. The chemistry of water reactor fuel

    International Nuclear Information System (INIS)

    Potter, P.E.

    1990-01-01

    In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

  19. Application of autoregressive methods and Lyapunov coefficients for instability studies of nuclear reactors; Aplicação de métodos autorregressivos e coeficientes de Lyapunov para estudos de instabilidades em reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Aruquipa Coloma, Wilmer

    2017-07-01

    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

  20. Proposed method of the modeling and simulation of corrosion product behavior in the primary cooling system of fast breeder reactors

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    2011-01-01

    Radioactive corrosion products (CP) are main cause of personal radiation exposure during maintenance without fuel failure in FBR plants. In order to establish the techniques of radiation dose estimation for worker in radiation-controlled area, Program SYstem for Corrosion Hazard Evaluation code 'PSYCHE' has been developed. The PSYCHE is based on the Solution-Precipitation model. The CP transfer calculation using the Solution-Precipitation model needs a fitting factor for the calculation of the precipitation of CP. This fitting factor must be determined based on the measured values in reactors that have operating experience. For this reason, the inability to make accurate predictions for reactor without measured values is a major issue. In this study, in addition to existing Solution-Precipitation model in PSYCHE, a transfer-model of CP species in particle form was applied to calculations of CP behavior in the primary cooling system of fast breeder reactor MONJU. Based on the calculated results, we estimated the contribution of CP deposition in the particle-form. It was suggested that the improved model including transfer-model of CP species in particle-form could be used for evaluation of CP transfer and radiation-source distribution in place of conventional Solution-Precipitation model with fitting factor in the PSYCHE. Moreover, it was predicted that CP particles would tend to be deposited in region with high-flow rate of coolant. (author)

  1. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  2. Effective search for stable segregation configurations at grain boundaries with data-mining techniques

    Science.gov (United States)

    Kiyohara, Shin; Mizoguchi, Teruyasu

    2018-03-01

    Grain boundary segregation of dopants plays a crucial role in materials properties. To investigate the dopant segregation behavior at the grain boundary, an enormous number of combinations have to be considered in the segregation of multiple dopants at the complex grain boundary structures. Here, two data mining techniques, the random-forests regression and the genetic algorithm, were applied to determine stable segregation sites at grain boundaries efficiently. Using the random-forests method, a predictive model was constructed from 2% of the segregation configurations and it has been shown that this model could determine the stable segregation configurations. Furthermore, the genetic algorithm also successfully determined the most stable segregation configuration with great efficiency. We demonstrate that these approaches are quite effective to investigate the dopant segregation behaviors at grain boundaries.

  3. Application of noise analysis on the operation of the RSG-GAS reactor

    International Nuclear Information System (INIS)

    Surian P, Tukiran S

    1999-01-01

    The RSG-GAS reactor has been operating for radioisotope production and experiments so that it is necessary to perform analysis of the reactor operation. analysis was done based on reactor noise experiment. Neutron noise at low and high power of the RSG-GAS has been analyzed using time and frequency domain with aim to determine the safety of reactor operation. The safety of reactor operation based on two parameters as, prompt neutron decay constant and decay ratio. The parameters are useful for reactor operation, so it is necessary to determine accurately. For determining prompt neutron decay constant, neutron density in the core of reactor which operated at 10 k W, was collected by using Fission Chamber detectors (FC). Based on power spectral density (PSD) was achieved break frequency about 23 Hz, so that the prompt neutron decay constant is about 151 sec -1 . While at at high power 20 MW, neutron density was collected by using Compensated Ionization Chamber (CIC) detector. The result at high power showed that there is reactivity effect in the core because of fluctuation in temperature and density of the coolant, and the decay ratio of 0.20, showed that the reactor is still operation in stable

  4. CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

    International Nuclear Information System (INIS)

    Nassersharif, B.; Peer, J.S.; DeHart, M.D.

    1987-01-01

    Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

  5. 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.

  6. 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

  7. Improving biogas production from anaerobic co-digestion of Thickened Waste Activated Sludge (TWAS) and fat, oil and grease (FOG) using a dual-stage hyper-thermophilic/thermophilic semi-continuous reactor.

    Science.gov (United States)

    Alqaralleh, Rania Mona; Kennedy, Kevin; Delatolla, Robert

    2018-07-01

    This paper investigates the feasibility and advantages of using a dual-stage hyper-thermophilic/thermophilic semi-continuous reactor system for the co-digestion of Thickened Waste Activated Sludge (TWAS) and Fat, Oil and Grease (FOG) to produce biogas in high quantity and quality. The performance of the dual-stage hyper-thermophilic (70°C)/thermophilic (55°C) anaerobic co-digestion system is evaluated and compared to the performance of a single-stage thermophilic (55°C) reactor that was used to co-digest the same FOG-TWAS mixtures. Both co-digestion reactors were compared to a control reactor (the control reactor was a single-stage thermophilic reactor that only digested TWAS). The effect of FOG% in the co-digestion mixture (based on total volatile solids) and the reactor hydraulic retention time (HRT) on the biogas/methane production and the reactors' performance were thoroughly investigated. The FOG% that led to the maximum methane yield with a stable reactor performance was determined for both reactors. The maximum FOG% obtained for the single-stage thermophilic reactor at 15 days HRT was found to be 65%. This 65% FOG resulted in 88.3% higher methane yield compared to the control reactor. However, the dual-stage hyper-thermophilic/thermophilic co-digestion reactor proved to be more efficient than the single-stage thermophilic co-digestion reactor, as it was able to digest up to 70% FOG with a stable reactor performance. The 70% FOG in the co-digestion mixture resulted in 148.2% higher methane yield compared to the control at 15 days HRT. 70% FOG (based on total volatile solids) is so far the highest FOG% that has been proved to be useful and safe for semi-continuous reactor application in the open literature. Finally, the dual-stage hyper-thermophilic/thermophilic co-digestion reactor also proved to be efficient and stable in co-digesting 40% FOG mixtures at lower HRTs (i.e., 9 and 12 days) and still produce high methane yields and Class A effluents

  8. The influence Of On Power Target Loading At RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Kusno; Pardi

    2001-01-01

    One of utilization purposes of the RSG-GAS reactor is to produce radioisotope. The target insertion or withdrawal on the irradiation position D-6 and E-7 in the core can be done manually, during reactor operation or shut down condition. The problem arises when the loading is under operation mode, because of flow and reactivity change the reactivity. The behavior of operation parameter changes due to target handling will be investigated in this paper. The behavior will be applied as a guidance of the operating group in handling the irradiation target

  9. Perturbation analysis of the TRIGA Mark II reactor Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)

    2013-04-15

    The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

  10. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    International Nuclear Information System (INIS)

    Wright, A.E.; Dutt, D.S.; Harrison, L.J.

    1990-01-01

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

  11. 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...

  12. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  13. 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.

  14. Design study of 'HIBLIC-I' reactor cavity

    International Nuclear Information System (INIS)

    Fujiie, Y.

    1984-01-01

    A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

  15. Chemical-looping combustion in a reverse-flow fixed bed reactor

    International Nuclear Information System (INIS)

    Han, Lu; Bollas, George M.

    2016-01-01

    A reverse-flow fixed bed reactor concept for CLC (chemical-looping combustion) is explored. The limitations of conventional fixed bed reactors, as applied to CLC, are overcome by reversing the gas flow direction periodically to enhance the mixing characteristics of the bed, thus improving oxygen carrier utilization and energy efficiency with respect to power generation. The reverse-flow reactor is simulated by a dusty-gas model and compared with an equivalent fixed bed reactor without flow reversal. Dynamic optimization is used to calculate conditions at which each reactor operates at maximum energy efficiency. Several cases studies illustrate the benefits of reverse-flow operation for the CLC with CuO and NiO oxygen carriers and methane and syngas fuels. The results show that periodic reversal of the flow during reduction improves the contact between the fuel and unconverted oxygen carrier, enabling the system to suppress unwanted catalytic reactions and axial temperature and conversion gradients. The operational scheme presented reduces the fluctuations of temperature during oxidation and increases the high-temperature heat produced by the process. CLC in a reverse-flow reactor has the potential to achieve higher energy efficiency than conventional fixed bed CLC reactors, when integrated with a downstream gas turbine of a combined cycle power plant. - Highlights: • Reverse-flow fixed bed CLC reactors for combined cycle power systems. • Dynamic optimization tunes operation of batch and transient CLC systems. • The reverse-flow CLC system provides stable turbine-ready gas stream. • Reverse-flow CLC fixed bed reactor has superior CO 2 capture and thermal efficiency.

  16. Effects of pyriproxyfen and buprofezin on immature development, female oviposition, and egg-hatching in the stable fly

    Science.gov (United States)

    The stable fly, Stomoxys calcitrans (L.) (Diptera: Muscidae), is one of the most economically significant biting flies affecting cattle. Use of traditional insecticides have only limited success in control of stable flies largely due to the stable fly’s unique feeding behaviors and immature developm...

  17. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  18. Dynamic modeling of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    March-Leuba, J.; Ibn-Khayat, M.

    1990-01-01

    The purpose of this paper is to provide a summary description and some applications of a computer model that has been developed to simulate the dynamic behavior of the advanced neutron source (ANS) reactor. The ANS dynamic model is coded in the advanced continuous simulation language (ACSL), and it represents the reactor core, vessel, primary cooling system, and secondary cooling systems. The use of a simple dynamic model in the early stages of the reactor design has proven very valuable not only in the development of the control and plant protection system but also of components such as pumps and heat exchangers that are usually sized based on steady-state calculations

  19. Digestion of cheese whey with anaerobic rotating biological contact reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lo, K V; Liao, P H

    1986-01-01

    A laboratory-scale anaerobic rotating biological contact reactor receiving full strength cheese whey was studied over a range of hydraulic retention times from 11 to 5 days at 35 degrees C. Methane production rates ranging from 1.68 to 3.26 litres CH/sub 4//litre/day and a 76 to 93% reduction in chemical oxygen demand were achieved. At hydraulic retention times shorter than 5 days, steady-state operation could not be maintained for reactors receiving either full strength or diluted whey. A two-stage fermentation system was also studied; the results indicated that stable operation and treatment efficiency (89.5% COD removal) could be achieved.

  20. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  1. Development of a novel integrated continuous reactor system for biocatalytic production of biodiesel.

    Science.gov (United States)

    Chattopadhyay, Soham; Sen, Ramkrishna

    2013-11-01

    A novel integrated immobilized enzyme-reactor system involving a continuous stirred tank reactor with two packed bed reactors in series was developed for the continuous production of biodiesel. The problem of methanol solubility into oil was solved by introducing a stirred tank reactor to dissolve methanol into partially converted oil. This step made the process perfectly continuous without requiring any organic solvent and intermittent methanol addition in the process. The substrate feeding rate of 0.74 mL/min and enzyme loading of 0.75 g per reactor were determined to be optimum for maximum biodiesel yield. The integrated continuous process was stable up to 45 cycles with biodiesel productivity of 137.2 g/L/h, which was approximately 5 times higher than solvent free batch process. In comparison with the processes reported in literature using expensive Novozyme 435 and hazardous organic solvent, the present process is completely green and perfectly continuous with economic and environmental advantages. Copyright © 2013 Elsevier Ltd. All rights reserved.

  2. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    Lundberg, L.B.; Croson, M.L.

    1994-01-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  3. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  4. Determination of the neutron flux for a possible way of controlling a fast reactor through the reflector

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1979-08-01

    The determination of time dependent flux is made in a fast reactor with the core embraced by a perfect reflector. The fuel burnup is taken in account establishing a nonlinear diffusion problem. A stable numeric scheme is done and the integration of two limit cases is obtained. Finally, one possibility of reactor control through the variation between two cases is discussed. (E.G.) [pt

  5. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

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

  6. Guidelines on the use of stable iodine as a prophylactic measure during nuclear emergencies

    International Nuclear Information System (INIS)

    1995-09-01

    Among the fission products that may be released in a power reactor accident the radio iodines are unique in that the dose received by persons exposed by inhalation of radio iodines in the gaseous plume may be substantially reduced. GMA-9 provides guidance on the medical aspects of the use of stable iodine compounds as a prophylactic measure in the event of a nuclear accident. A review of the physiologic basis for the use of stable iodine as a prophylactic measure and the effects of radiation on the thyroid gland are provided. Logistic factors that must be considered to provide the optimum level of radiological protection and medical safety are also addressed. 71 refs., 6 tabs., 2 figs

  7. Guidelines on the use of stable iodine as a prophylactic measure during nuclear emergencies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Among the fission products that may be released in a power reactor accident the radio iodines are unique in that the dose received by persons exposed by inhalation of radio iodines in the gaseous plume may be substantially reduced. GMA-9 provides guidance on the medical aspects of the use of stable iodine compounds as a prophylactic measure in the event of a nuclear accident. A review of the physiologic basis for the use of stable iodine as a prophylactic measure and the effects of radiation on the thyroid gland are provided. Logistic factors that must be considered to provide the optimum level of radiological protection and medical safety are also addressed. 71 refs., 6 tabs., 2 figs.

  8. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

    A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

  9. Nitrate removal from high strength nitrate-bearing wastes in granular sludge sequencing batch reactors.

    Science.gov (United States)

    Krishna Mohan, Tulasi Venkata; Renu, Kadali; Nancharaiah, Yarlagadda Venkata; Satya Sai, Pedapati Murali; Venugopalan, Vayalam Purath

    2016-02-01

    A 6-L sequencing batch reactor (SBR) was operated for development of granular sludge capable of denitrification of high strength nitrates. Complete and stable denitrification of up to 5420 mg L(-1) nitrate-N (2710 mg L(-1) nitrate-N in reactor) was achieved by feeding simulated nitrate waste at a C/N ratio of 3. Compact and dense denitrifying granular sludge with relatively stable microbial community was developed during reactor operation. Accumulation of large amounts of nitrite due to incomplete denitrification occurred when the SBR was fed with 5420 mg L(-1) NO3-N at a C/N ratio of 2. Complete denitrification could not be achieved at this C/N ratio, even after one week of reactor operation as the nitrite levels continued to accumulate. In order to improve denitrification performance, the reactor was fed with nitrate concentrations of 1354 mg L(-1), while keeping C/N ratio at 2. Subsequently, nitrate concentration in the feed was increased in a step-wise manner to establish complete denitrification of 5420 mg L(-1) NO3-N at a C/N ratio of 2. The results show that substrate concentration plays an important role in denitrification of high strength nitrate by influencing nitrite accumulation. Complete denitrification of high strength nitrates can be achieved at lower substrate concentrations, by an appropriate acclimatization strategy. Copyright © 2015 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  10. Uniaxial low cycle fatigue behavior for pre-corroded 16MND5 bainitic steel in simulated pressurized water reactor environment

    Science.gov (United States)

    Chen, Xu; Ren, Bin; Yu, Dunji; Xu, Bin; Zhang, Zhe; Chen, Gang

    2018-06-01

    The effects of uniaxial tension properties and low cycle fatigue behavior of 16MND5 bainitic steel cylinder pre-corroded in simulated pressurized water reactor (PWR) were investigated by fatigue at room temperature in air and immersion test system, scanning electron microscopy (SEM), energy disperse spectroscopy (EDS). The experimental results indicated that the corrosion fatigue lives of 16MND5 specimen were significantly affected by the strain amplitude and simulated PWR environments. The compositions of corrosion products were complexly formed in simulated PWR environments. The porous corrosion surface of pre-corroded materials tended to generate pits as a result of promoting contact area to the fresh metal, which promoted crack initiation. For original materials, the fatigue cracks initiated at inclusions imbedded in the micro-cracks. Moreover, the simulated PWR environments degraded the mechanical properties and low cycle fatigue behavior of 16MND5 specimens remarkably. Pre-corrosion of 16MND5 specimen mainly affected the plastic term of the Coffin-Manson equation.

  11. The behavior of shallow flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Rolfe, S.T.

    1991-11-01

    Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs

  12. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  13. Project in fiscal 1988 for research and development of basic technologies in next generation industries. Achievement report on research and development of bio-reactors; 1988 nendo bio reactor no kenkyu kaihatsu seika hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-03-01

    Research and development has been performed on bio-reactors to substitute the oxidizing reaction process, and on reduction reacting bio-reactors to fix the coenzyme regeneration system. This paper summarizes the achievements in fiscal 1988. In the research of a bio-reactor using bacteria that produce muconic acid from benzoic acid, stable productivity of 60 g/L/day or more was obtained. In the research of a multi-phase based bio-reactor composed of air, oil, water and biomass, discussions were given on the phase inverting film permeation type reactor. In the research of a bio-reactor to produce acetic acid from carbon dioxide and hydrogen, productivity of 149 g/L/day as maximum was achieved by enhancing the production speed by means of pressurization. In the research of a bio-reactor to produce hydroquinone from phenol, up-keeping the duration for 100 hours or longer has become possible at the hydroquinone production speed of 3 g/L/h. In the research of a reduction-based bio-reactor incorporating the regeneration system of coenzyme NAD(P)H, discussions were given on optimizing the continuous enzymatic reaction in the production of sorbitol. (NEDO)

  14. In-reactor fatigue crack propagation

    International Nuclear Information System (INIS)

    Ermi, A.M.; Mervyn, D.A.; Straalsund, J.L.

    1979-08-01

    An in-reactor fatigue experiment is being designed to determine the effect of dynamic irradiation on the fatigue crack propagation (FCP) behavior of candidate fusion first wall materials. This investigation has been prompted by studies which show gross differences in crack growth characteristics of creep rupture specimens testing by postirradiation versus dynamic in-reactor methods. The experiment utilizes miniature center-cracked-tension specimens developed specifically for in-reactor studies. In the test, a chain of eight specimens, precracked to various initial crack lengths, is stressed during irradiation to determine crack growth rate as a function of stress intensity. Load levels were chosen which result in small crack growth rates encompassing a regime of the crack growth curve not previously investigated during irradiation studies of FCP. The test will be conducted on 20% cold worked 316 stainless steel at a temperature of 425 0 C, in a sodium environment, and at a frequency of 1 cycle/min. Irradiation will occur in the Oak Ridge Research Reactor, resulting in a He/dpa ratio similar to that expected at the first wall in a fusion reactor. Detailed design of the experiment is presented, along with crack growth data obtained from prototypic testing of the experimental apparatus. These results are compared to data obtained under similar conditions generated by conventional test methods

  15. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

  16. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  17. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

    2010-01-31

    An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  18. Education and training by utilizing irradiation test reactor simulator

    International Nuclear Information System (INIS)

    Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi

    2016-01-01

    The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)

  19. A Green Process for High-Concentration Ethylene and Hydrogen Production from Methane in a Plasma-Followed-by-Catalyst Reactor

    International Nuclear Information System (INIS)

    Wang Kangjun; Li Xiaosong; Zhu Aimin

    2011-01-01

    A green process for the oxygen-free conversion of methane to high-concentration ethylene and hydrogen in a plasma-followed-by-catalyst (PFC) reactor is presented. Without any catalysts and with pure methane used as the feed gas, a stable kilohertz spark discharge leads to an acetylene yield of 64.1%, ethylene yield of 2.5% and hydrogen yield of 59.0% with 80.0% of methane conversion at a methane flow rate of 50 cm 3 /min and a specific input energy of 38.4 kJ/L. In the effluent gas from a stable kilohertz spark discharge reactor, the concentrations of acetylene, ethylene and hydrogen were 18.1%, 0.7% and 66.9%, respectively. When catalysts Pd-Ag/SiO 2 were employed in the second stage with discharge conditions same as in the case of plasma alone, the PFC reactor provides an ethylene yield of 52.1% and hydrogen yield of 43.4%. The concentrations of ethylene and hydrogen in the effluent gas from the PFC reactor were found to be as high as 17.1% and 62.6%, respectively. Moreover, no acetylene was detected in the effluent gas. This means that a high concentration of ethylene and oxygen-free hydrogen can be co-produced directly from methane in the PFC reactor.

  20. Thermohydraulic accident behavior of reactors. Thermohydraulisches Stoerfallverhalten der Reaktoren in Greifswald

    Energy Technology Data Exchange (ETDEWEB)

    Horche, W.; Kirmse, R.; Reichenbach, D.; Weber, J.P. (Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany))

    1992-12-01

    GRS, on behalf of the German Federal Ministry for the Environment, conducted an assessment of the technical safety of the Greifswald nuclear generating units of the Soviet WWER-440/W-230 and W-213 reactor lines, respectively. The evaluation of existing accident analyses and the execution of some first calculations by GRS added to the know-how of GRS. This is reflected in the increased participation by GRS in international expert bodies investigating safety problems of WWER-440 plants. The contributions made towards international WWER projects within the framework of IAEA missions or as a result of bilateral consultations strengthen international partnership in the field of reactor safety in Central and Eastern Europe. (orig.).

  1. Conceptual design for simulator of irradiation test reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

    2012-03-01

    A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

  2. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  3. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  4. Behavior of generated aerosols in decommissioning of reactor

    International Nuclear Information System (INIS)

    Tomii, H.; Nakamura, K.

    1999-01-01

    Generated aerosols in dismantling of the JPDR were investigated for making an estimation of air contamination. The maximum dose equivalent rate at the surface of each reactor component was 9.4 Sv/h for core shroud, 80 mSv/h for pressure vessel, 2.0 mSv/h for biological shield, respectively. An under-water cutting method with remote handling plasma torch was used for dismantling of the core shroud and the pressure vessel. The biological shield was dismantled by an in-air cutting method and a controlled blasting method. Pipes connected to recirculation system were dismounted by a conventional mechanical and thermal cutting machine in the air. Generated radioactive aerosols were collected in the exhaust air of green house which enclosed the upper part of the reactor room to control the air contamination. An Andersen sampler was used for the measurement of particle distribution in the aerosols. Most of the particle size was below 0.1 μm in the under-water cutting method. The particle size distribution in the in-air cutting method, however, was divided into two parts at 0.1 μm and 0.3 μm. Dispersion rate of aerosol into the atmosphere was decreased exponentially with the depth of water. The dispersion rate and the size distribution of aerosol generated during cutting of the stainless steel pipes and blasting of the biological shield are also reported in the paper. (Suetake, M.)

  5. High precision and stable structures for particle detectors

    CERN Document Server

    Da Mota Silva, S; Hauviller, Claude

    1999-01-01

    The central detectors used in High Energy Physics Experiments require the use of light and stable structures capable of supporting delicate and precise radiation detection elements. These structures need to be highly stable under environmental conditions where external vibrations, high radiation levels, temperature and humidity gradients should be taken into account. Their main design drivers are high dimension and dynamic stability, high stiffness to mass ratio and large radiation length. For most applications, these constraints lead us to choose Carbon Fiber Reinforced Plastics ( CFRP) as structural element. The construction of light and stable structures with CFRP for these applications can be achieved by careful design engineering and further confirmation at the prototyping phase. However, the experimental environment can influence their characteristics and behavior. In this case, theuse of adaptive structures could become a solution for this problem. We are studying structures in CFRP with bonded piezoel...

  6. Experimental and analytical studies on soil-structure interaction behavior of nuclear reactor building

    International Nuclear Information System (INIS)

    Tsushima, Y.

    1978-01-01

    The purpose of this study is to estimate damping effects due to soil-structure interaction by the dissipation of vibrational energy to the ground through the foundation in a building with a short fundamental period such as a nuclear reactor building. The author performed experimental and analytical studies on the vibrational characteristics of model steel structures ranging from one to four stories high erected on the rigid base and located on soil, which are simulated from the vibrational characteristics of a prototype reactor building: the former study is to obtain damping effects due to inner friction of steel frames and the latter to obtain radiation damping effects due to soil-structure interaction. The author also touches upon the results of experiments performed on a BWR-type reactor building in 1974, which showed damping ratios higher than 20% of those in fundamental modes. Then the author attempts to estimate the damping effects of the reactor building by his own method proposed in the report. Through these studies the author finally concludes that the experimental damping effects are remarkable in the lower modes by the energy dissipation and the analytical results show a fairly good fit to the experimental ones

  7. Device for controlling water supply to nuclear reactor

    International Nuclear Information System (INIS)

    Iwasaki, Toshio.

    1974-01-01

    Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

  8. Materials behavior in interim storage of spent fuel

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Gilbert, E.R.; Inman, S.C.

    1982-01-01

    Interim storage has emerged as the only current spent-fuel management method in the US and is essential in all countries with nuclear reactors. Materials behavior is a key aspect in licensing interim-storage facilities for several decades of spent-fuel storage. This paper reviews materials behavior in wet storage, which is licensed for light-water reactor (LWR) fuel, and dry storage, for which a licensing position for LWR fuel is developing

  9. 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

  10. Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

    International Nuclear Information System (INIS)

    1978-07-01

    The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

  11. Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

    2009-01-01

    The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

  12. The tempered stable process with infinitely divisible inverse subordinators

    International Nuclear Information System (INIS)

    Wyłomańska, Agnieszka

    2013-01-01

    In the last decade processes driven by inverse subordinators have become extremely popular. They have been used in many different applications, especially for data with observable constant time periods. However, the classical model, i.e. the subordinated Brownian motion, can be inappropriate for the description of observed phenomena that exhibit behavior not adequate for Gaussian systems. Therefore, in this paper we extend the classical approach and replace the Brownian motion by the tempered stable process. Moreover, on the other hand, as an extension of the classical model, we analyze the general class of inverse subordinators. We examine the main properties of the tempered stable process driven by inverse subordinators from the infinitely divisible class of distributions. We show the fractional Fokker–Planck equation of the examined process and the asymptotic behavior of the mean square displacement for two cases of subordinators. Additionally, we examine how an external force can influence the examined characteristics. (paper)

  13. Treatment of the response of a reactor to stochastic reactivity input

    International Nuclear Information System (INIS)

    Bansal, N.K.

    1977-08-01

    One of the important applications of reactor noise theory, which relies on the methematical methods for treating stochastic processes, is to determine either the confidence limits for the allowed deviations of the measured signals during normal reactor operation, or the statistical properties of their respective expectation values. In this report, we stress mainly the general mathematical aspects for treating this problem. A global description of a reactor system, perturbed by stochastic reactivity input, leads to a stochastic differential equation with parametric excitation. A discrepancy exists in literature about obtaining the correct solution of such an equation in its general frame. We discuss this discrepancy and review the work done for solving such an equation. Some recent work indicates that linearisation of system's equations is justified in most cases of reactor operations. We develop a general scheme for calculating the various covariances and correlation functions in a stable and stationary system, which is perturbed by various noise sources and where linearisation of system's equations is justified. The formulation is easily extendable to an unstable, nonstationary system, like an uncontrolled critical reactor as demonstrated. (orig.) [de

  14. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  15. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling.

  16. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    International Nuclear Information System (INIS)

    Lee, Choong Wie; Kim, Hee Reyoung

    2015-01-01

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling

  17. Collection of summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1995

    International Nuclear Information System (INIS)

    1996-07-01

    This report meeting was held on May 22, 1995 at University of Tokyo by about 40 participants. As the topics on the fusion reactor engineering research in Japan, lectures were given on the present state and future of nuclear fusion networks and on the strong magnetic field tokamak using electromagnetic force-balanced coils being planned. Thereafter, the reports of the results of the researches which were carried out by using this experimental facility were made, centering around the subject related to the future conception 'The interface properties of fusion reactor materials and particle transport control'. The publication was made on the future conception of the basic experiment setup for fusion reactor blanket design, the application of high temperature superconductors to the advancement of nuclear fusion reactors, the modeling of the dynamic irradiation behavior of fusion reactor materials, the interface particle behavior in plasma-wall interaction, the behavior of tritium on the surface of breeding materials, and breeding materials and the behavior of tritium in plasma-wall interaction. (K.I.)

  18. Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

    International Nuclear Information System (INIS)

    Chen Daolong

    1986-01-01

    A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

  19. Transferability of results of PTS experiments to the integrity assessment of reactor pressure vessels

    International Nuclear Information System (INIS)

    Roos, E.; Eisele, U.; Stumpfrock, L.

    1997-01-01

    The integrity assessment of the reactor pressure vessel (RPV) is based on the fracture mechanics concept as provided in the code. However this concept covers only the linear-elastic fracture mechanics regime on the basis of the reference temperature RT NDT as derived from charpy impact and drop-weight test. The conservatism of this concept was demonstrated for a variety of different materials covering optimized and lower bound material states with regard to unirradiated and irradiated conditions. For the elastic-plastic regime, methodologies have been developed to describe ductile crack initiation and stable crack growth. The transferability of both, the linear-elastic and elastic-plastic fracture mechanics concept was investigated with the help of large scale specimens focusing on complex loading situations as they result from postulated thermal shock events for the RPV. A series of pressurized thermal shock (PTS) experiments were performed in which the applicability of the fracture mechanics parameters derived from small scale specimen testing could be demonstrated. This includes brittle (static and dynamic) crack initiation and crack arrest in the low charpy energy regime as well as stable crack initiation, stable crack growth and crack arrest in the upper shelf toughness regime. The paper provides the basic material data, the load paths, representative for large complex components as well as experimental and theoretical results of PTS experiments. From these data it can be concluded that the available fracture mechanics concepts can be used to describe the component behavior under transient loading conditions. (author). 26 refs, 12 figs, 1 tab

  20. Transferability of results of PTS experiments to the integrity assessment of reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Roos, E; Eisele, U; Stumpfrock, L [MPA Stuttgart (Germany)

    1997-09-01

    The integrity assessment of the reactor pressure vessel (RPV) is based on the fracture mechanics concept as provided in the code. However this concept covers only the linear-elastic fracture mechanics regime on the basis of the reference temperature RT{sub NDT} as derived from charpy impact and drop-weight test. The conservatism of this concept was demonstrated for a variety of different materials covering optimized and lower bound material states with regard to unirradiated and irradiated conditions. For the elastic-plastic regime, methodologies have been developed to describe ductile crack initiation and stable crack growth. The transferability of both, the linear-elastic and elastic-plastic fracture mechanics concept was investigated with the help of large scale specimens focusing on complex loading situations as they result from postulated thermal shock events for the RPV. A series of pressurized thermal shock (PTS) experiments were performed in which the applicability of the fracture mechanics parameters derived from small scale specimen testing could be demonstrated. This includes brittle (static and dynamic) crack initiation and crack arrest in the low charpy energy regime as well as stable crack initiation, stable crack growth and crack arrest in the upper shelf toughness regime. The paper provides the basic material data, the load paths, representative for large complex components as well as experimental and theoretical results of PTS experiments. From these data it can be concluded that the available fracture mechanics concepts can be used to describe the component behavior under transient loading conditions. (author). 26 refs, 12 figs, 1 tab.

  1. Method of controlling the reactor operation

    International Nuclear Information System (INIS)

    Ishiguro, Akira; Nakakura, Hiroyuki.

    1987-01-01

    Purpose: To moderate vibratory response due to delayed operation thereby obtain stable controlled response in the operation control for a PWR type reactor. Method: the reactor operation is controlled by the axial power distribution control by regulating the boron concentration in primary coolants with a boron density control system and controlling the average temperature for the primary coolants with the control rod control system. In this case, the control operation and the control response become instable due to transmission delay, etc. of aqueous boric acid injection to the primary coolant circuits to result in vibratory response. In the present invention, signals are prepared by adding the amount in proportion to the variation coefficient with time of xenone concentration obtained from the measured value for the reactor power added to the conventional axial power distribution parameter deviation and used as the input signals for the boron concentration control system. As a result, the instability due to the transmission delay of the aqueous boric acid injection is improved by the preceding control by the amount in proportion with the variation coefficient with time of the xenone concentration. An advantageous effect can be expected for the load following operation during day time according to the present invention. (Kamimura, M.)

  2. Low temperature catalytic combustion of propane over Pt-based catalyst with inverse opal microstructure in microchannel reactor

    NARCIS (Netherlands)

    Guan, G.; Zapf, R.; Kolb, G.A.; Men, Y.; Hessel, V.; Löwe, H.; Ye, J.; Zentel, R.

    2007-01-01

    novel Pt-based catalyst with highly regular, periodic inverse opal microstructure was fabricated in a microchannel reactor, and catalytic testing revealed excellent conversion and stable activity for propane combustion at low temperatures

  3. Stable thermophilic anaerobic digestion of dissolved air flotation (DAF) sludge by co-digestion with swine manure.

    Science.gov (United States)

    Creamer, K S; Chen, Y; Williams, C M; Cheng, J J

    2010-05-01

    Environmentally sound treatment of by-products in a value-adding process is an ongoing challenge in animal agriculture. The sludge produced as a result of the dissolved air flotation (DAF) wastewater treatment process in swine processing facilities is one such low-value residue. The objective of this study was to determine the fundamental performance parameters for thermophilic anaerobic digestion of DAF sludge. Testing in a semi-continuous stirred tank reactor and in batch reactors was conducted to determine the kinetics of degradation and biogas yield. Stable operation could not be achieved using pure DAF sludge as a substrate, possibly due to inhibition by long-chain fatty acids or to nutrient deficiencies. However, in a 1:1 ratio (w/w, dry basis) with swine manure, operation was both stable and productive. In the semi-continuous stirred reactor at 54.5 degrees Celsius, a hydraulic residence time of 10 days, and an organic loading rate of 4.68 gVS/day/L, the methane production rate was 2.19 L/L/day and the specific methane production rate was 0.47 L/gVS (fed). Maximum specific methanogenic activity (SMA) in batch testing was 0.15 mmoles CH(4) h(-1) gVS(-1) at a substrate concentration of 6.9 gVS L(-1). Higher substrate concentrations cause an initial lag in methane production, possibly due to long-chain fatty acid or nitrogen inhibition. Copyright 2009 Elsevier Ltd. All rights reserved.

  4. Fiscal year 1999 multi-year work plan, advanced reactors transition program

    International Nuclear Information System (INIS)

    Gantt, D.A.

    1998-01-01

    The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart

  5. Fiscal year 1999 multi-year work plan, advanced reactors transition program

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1998-09-17

    The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart.

  6. The Canadian development program for conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    Charlesworth, D.H.; Bourns, W.T.; Buckley, L.P.

    1978-07-01

    Currently, radioactive wastes arising from the operation of Canadian nuclear reactors are placed in interim storage in concrete containment structures except for gaseous and liquid wastes containing small amounts of radioactivity which are dispersed. With the objective of replacing storage by permanent disposal, a program is underway to develop and demonstrate an integrated process for converting all reactor wastes to a stable, leach-resistant form which will immobilize the radionuclides in the waste repository. The major tool for this development is a Waste Treatment Centre, now being constructed at Chalk River Nuclear Laboratories, which will combine reverse-osmosis, incineration, evaporation and bituminizing processes. (author)

  7. Paleoseismicity of two historically quiescent faults in Australia: Implications for fault behavior in stable continental regions

    Science.gov (United States)

    Crone, A.J.; De Martini, P. M.; Machette, M.M.; Okumura, K.; Prescott, J.R.

    2003-01-01

    Paleoseismic studies of two historically aseismic Quaternary faults in Australia confirm that cratonic faults in stable continental regions (SCR) typically have a long-term behavior characterized by episodes of activity separated by quiescent intervals of at least 10,000 and commonly 100,000 years or more. Studies of the approximately 30-km-long Roopena fault in South Australia and the approximately 30-km-long Hyden fault in Western Australia document multiple Quaternary surface-faulting events that are unevenly spaced in time. The episodic clustering of events on cratonic SCR faults may be related to temporal fluctuations of fault-zone fluid pore pressures in a volume of strained crust. The long-term slip rate on cratonic SCR faults is extremely low, so the geomorphic expression of many cratonic SCR faults is subtle, and scarps may be difficult to detect because they are poorly preserved. Both the Roopena and Hyden faults are in areas of limited or no significant seismicity; these and other faults that we have studied indicate that many potentially hazardous SCR faults cannot be recognized solely on the basis of instrumental data or historical earthquakes. Although cratonic SCR faults may appear to be nonhazardous because they have been historically aseismic, those that are favorably oriented for movement in the current stress field can and have produced unexpected damaging earthquakes. Paleoseismic studies of modern and prehistoric SCR faulting events provide the basis for understanding of the long-term behavior of these faults and ultimately contribute to better seismic-hazard assessments.

  8. A Compact Unconditionally Stable Method for Time-Domain Maxwell's Equations

    Directory of Open Access Journals (Sweden)

    Zhuo Su

    2013-01-01

    Full Text Available Higher order unconditionally stable methods are effective ways for simulating field behaviors of electromagnetic problems since they are free of Courant-Friedrich-Levy conditions. The development of accurate schemes with less computational expenditure is desirable. A compact fourth-order split-step unconditionally-stable finite-difference time-domain method (C4OSS-FDTD is proposed in this paper. This method is based on a four-step splitting form in time which is constructed by symmetric operator and uniform splitting. The introduction of spatial compact operator can further improve its performance. Analyses of stability and numerical dispersion are carried out. Compared with noncompact counterpart, the proposed method has reduced computational expenditure while keeping the same level of accuracy. Comparisons with other compact unconditionally-stable methods are provided. Numerical dispersion and anisotropy errors are shown to be lower than those of previous compact unconditionally-stable methods.

  9. Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2015-01-15

    Highlights: • Dislocation loops were the prominent defect, but neutron irradiation caused higher loop density. • Grain boundaries had similar amounts of radiation-induced segregation. • The increment in hardness and yield stress due to irradiation were very similar. • Relative IASCC susceptibility was nearly identical. • The effect of dislocation channel step height on IASCC was similar. - Abstract: The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni–Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed

  10. The feasibility of trace element supplementation for stable operation of wheat stillage-fed biogas tank reactors.

    Science.gov (United States)

    Gustavsson, J; Svensson, B H; Karlsson, A

    2011-01-01

    The aim of this study was to investigate the effect of trace element supplementation on operation of wheat stillage-fed biogas tank reactors. The stillage used was a residue from bio-ethanol production, containing high levels of sulfate. In biogas production, high sulfate content has been associated with poor process stability in terms of low methane production and accumulation of process intermediates. However, the results of the present study show that this problem can be overcome by trace element supplementations. Four lab-scale wheat stillage-fed biogas tank reactors were operated for 345 days at a hydraulic retention time of 20 days (37 degrees C). It was concluded that daily supplementation with Co (0.5 mg L(-1)), Ni (0.2 mg L(-1)) and Fe (0.5 g L(-1)) were required for maintaining process stability at the organic loading rate of 4.0 g volatile solids L(-1) day(-1).

  11. Thermal Behavior of the Coolant in the Emergency Cooldown Tank for an Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Joo Hyung; Kim, Seok; Kim, Woo Shik; Jung, Seo Yoon; Kim, Young In [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The Residual Heat Removal System (PRHRS) is one of the passive safety systems which should be activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators until the safe shutdown conditions are reached. In the previous study presented at the last KNS Autumn Meeting, transient behavior of the RCS temperature and the cooling performance of the PRHRS were investigated numerically by using newly developed in-house code based on MATLAB software. By using the program, the steady-state and transient (quasi-steady state) characteristics during the operation of the PRHRS had been reported. In this program, the temperature of the coolant in the Emergency Cooldown Tank (ECT) was assumed to be constant at saturated state and pool boiling heat transfer mechanism was applied through the entire time domain. The coolant of the ECT reached at a saturated state in early time. It was revealed that the assumption made in the previous study was reasonable.

  12. Experimental study on seismic behaviors of two-storied sophisticated model for nuclear reactor building

    Energy Technology Data Exchange (ETDEWEB)

    Higashiura, Akira; Sato, Kazuhide; Muramoto, Michiya; Yanagase, Takahito; Watanabe, Satoshi

    1987-03-01

    In this paper, by pseudo dynamic test using substructuring technique and lateral static loading test, authors wish to introduce the investigation on the seismic behaviors of nuclear reactor building. The results obtained by those test are as follows. 1) The maximum response displacements obtained by pseudo dynamic test are equivalent to those by dynamic analytical procedures using the approximate earthquake ground motion. 2) In the finally stage of pseudo dynamic test, the natural period of the system is increased about three times as long as that in elastic region. 3) Some shear cracks is observed on the web portion of the box and the truncated conical wall at the end of pseudo dynamic test. 4) Maximum shear forces in the test specimen obtained by pseudo dynamic test are about one third of the ultimate shear strength of it obtained by static loading test. 5) At the ultimate strength of the test specimen on static loading test, a lot of shear cracks and crush of concrete are observed on web portion of the box and the truncated conical wall.

  13. The promises and challenges of future reactor system developments

    International Nuclear Information System (INIS)

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

    2007-01-01

    Nuclear power is an inevitable option in Korea to overcome the scarcity of national energy resources and to reduce its overseas energy dependency. During the past three decades, Korea has accomplished outstanding achievements in facilitating a nuclear power development. The share of nuclear power in electricity generation has been rapidly increasing since 1978. Nuclear power has provided Korea with a most economically and environmentally-friendly way of generating electric energy, and has contributed a lot to its national economy growth. It will continue to do so in the future. For a stable and economical supply of electricity, nationwide efforts toward achieving self-reliance in nuclear power technology have been pursued. To date, a series of nuclear technology self-reliance programs such as CANDU fuel technology, PWR fuel technology, and nuclear reactor (KSNPP) technology have been successfully completed. KSNP is a technologically advanced power plant modified by Koreas' own operating experience and domestic technology and designed by adapting several advanced technologies suitable for its national situation. The KSNP was applied to the construction of Yonggwang 5 and 6 and Ulchin 5 and 6 and is now being replicated to provide a stable, economical and reliable electric power supply. Through a comprehensive nuclear Research and Development programs, an enhancement of its indigenous nuclear technology capability is currently being pursued. The effort has focused on improving its indigenous nuclear power technology such as improvements in safety and economy of the KSNP (KSNP+), a 600 MWe class KSNP and advanced fuels, and the establishment of industrial codes and standards. In addition, a Korean Advanced Power Reactor (APR 1400) and a System integrated Modular Advanced Reactor (SMART) are currently under development. The APR 1400 with a capacity of 1,400 MWe will be characterized by its drastically enhanced safety, reliability, and operability as well as its

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Proposal for a seismic facility for reactor safety research

    International Nuclear Information System (INIS)

    Anderson, C.A.; Dove, R.C.; Rhorer, R.L.

    1976-07-01

    Certain problem areas in the seismic analysis and design of nuclear reactors are enumerated and the way in which an experimental program might contribute to each area is examined. The use of seismic simulation testing receives particular attention, especially with regard to the verification of structural response analysis. The importance of scale modeling used in conjunction with seismic simulation is also stressed. The capabilities of existing seismic simulators are summarized, and a proposed facility is described which would considerably extend the ability to conduct, with confidence, confirmatory experiments on the behavior of reactor components when subjected to seismic excitation. Particular applications to gas-cooled and other reactor types are described

  16. Parametric Sensibility in Lixiviation Reactors

    Directory of Open Access Journals (Sweden)

    Dra. Margarita Rivera-Soto

    2015-11-01

    Full Text Available This work presents the results obtained in an analysis of the parametric sensibility, on the base of a mathematical model, which describes the behavior a lixiviation reactors battery inside the limits of the habitual work of the industrial plant, in a concrete process and of high complexity. The analysis was carried out with the purpose of determining the effect that the changes in different operation variables have on the behavior of the system and it gave as result that the most important variables are: the mineral-acid relationship, the concentration of magnesium and of nickel.

  17. Developmental Trajectories of Peer-Reported Aggressive Behavior: The Role of Friendship Understanding, Friendship Quality, and Friends' Aggressive Behavior.

    Science.gov (United States)

    Malti, Tina; McDonald, Kristina; Rubin, Kenneth H; Rose-Krasnor, Linda; Booth-LaForce, Cathryn

    2015-10-01

    To investigate developmental trajectories in peer-reported aggressive behavior across the transition from elementary-to-middle school, and whether aggressive behavior trajectories were associated with friendship quality, friends' aggressive behavior, and the ways in which children think about their friendships. Participants included a community sample of 230 5 th grade children who were assessed when they made a transition from elementary-to-middle school (6 th grade). Peer nominations were used to assess the target child's and friend's aggressive behavior. Self- and friend reports were used to measure friendship quality; friendship understanding was assessed via a structured interview. General Growth Mixture Modeling (GGMM) revealed three distinct trajectories of peer-reported aggressive behavior across the school transition: low-stable, decreasing, and increasing. Adolescents' understanding of friendship formation differentiated the decreasing from the low-stable aggressive behavior trajectories, and the understanding of friendship trust differentiated the increasing from the low-stable aggressive and decreasing aggressive behavior trajectories. The findings indicated that a sophisticated understanding of friendship may serve as a protective factor for initially aggressive adolescents as they transition into middle school. Promoting a deepened understanding of friendship relations and their role in one's own and others' well-being may serve as an important prevention and intervention strategy to reduce aggressive behavior.

  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. Iodine chemistry in a reactor regulation

    International Nuclear Information System (INIS)

    Powers, D.A.

    1996-01-01

    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

  20. Determining neutron multiplication factor in the infinite system by reactivity dependence on one dimension of the reactor core; Odredjivanje faktora umnozavanja neutrona u beskonacnom sistemu pomocu zavisnosti reaktivnosti od jedne dimenzije reaktorskog jezgra, Diplomski rad

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1975-07-01

    The objective of this task was to apply Fermi age theory for determining {tau} and neutron multiplication factor in infinite medium by measuring reactivity coefficient of heavy water in heterogeneous mixed reactor lattice. Basis of experiment is the measurement of stable reactor period. Measurement of heavy water reactivity coefficient by measuring the stable reactor period is described for chosen overcritical heavy water levels. Calculated values of infinite multiplication factor for measured neutron age data are presented and they are compared to expected theoretical values.

  1. The spatial kinetic analysis of accelerator-driven subcritical reactor

    International Nuclear Information System (INIS)

    Takahashi, H.; An, Y.; Chen, X.

    1998-02-01

    The operation of the accelerator driven reactor with subcritical condition provides a more flexible choice of the reactor materials and of design parameters. A deep subcriticality is chosen sometime from the analysis of point kinetics. When a large reactor is operated in deep subcritical condition by using a localized spallation source, the power distribution has strong spatial dependence, and point kinetics does not provide proper analysis for reactor safety. In order to analyze the spatial and energy dependent kinetic behavior in the subcritical reactor, the authors developed a computation code which is composed of two parts, the first one is for creating the group cross section and the second part solves the multi-group kinetic diffusion equations. The reactor parameters such as the cross section of fission, scattering, and energy transfer among the several energy groups and regions are calculated by using a code modified from the Monte Carlo codes MCNPA and LAHET instead of the usual analytical method of ANISN, TWOTRAN codes. Thus the complicated geometry of the accelerator driven reactor core can be precisely taken into account. The authors analyzed the subcritical minor actinide transmutor studied by Japan Atomic Energy Research Institute (JAERI) using the code

  2. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  3. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  4. Structure and creep of Russian reactor steels with a BCC structure

    Science.gov (United States)

    Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

    2017-05-01

    The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

  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. 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.

  7. Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

  8. Particle formation and its control in dual frequency plasma etching reactors

    International Nuclear Information System (INIS)

    Kim, Munsu; Cheong, Hee-Woon; Whang, Ki-Woong

    2015-01-01

    The behavior of a particle cloud in plasma etching reactors at the moment when radio frequency (RF) power changes, that is, turning off and transition steps, was observed using the laser-light-scattering method. Two types of reactors, dual-frequency capacitively coupled plasma (CCP) and the hybrid CCP/inductively coupled plasma (ICP), were set up for experiments. In the hybrid CCP/ICP reactor (hereafter ICP reactor), the position and shape of the cloud were strongly dependent on the RF frequency. The particle cloud becomes larger and approaches the electrode as the RF frequency increases. By turning the lower frequency power off later with a small delay time, the particle cloud is made to move away from the electrode. Maintaining lower frequency RF power only was also helpful to reduce the particle cloud size during this transition step. In the ICP reactor, a sufficient bias power is necessary to make a particle trap appear. A similar particle cloud to that in the CCP reactor was observed around the sheath region of the lower electrode. The authors can also use the low-frequency effect to move the particle cloud away from the substrate holder if two or more bias powers are applied to the substrate holder. The dependence of the particle behavior on the RF frequencies suggests that choosing the proper frequency at the right moment during RF power changes can reduce particle contamination effectively

  9. Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2012-10-01

    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∞) and the maximum power peaking factor (PPF max ), as well as the reactivity coefficients by the fuel temperature. The k∞ and PPF 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 2O3 ) in some rods of the same assembly. The obtained results show values k∞ and PPF 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)

  10. 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.

  11. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  12. Reduction of the hydraulic retention time at constant high organic loading rate to reach the microbial limits of anaerobic digestion in various reactor systems.

    Science.gov (United States)

    Ziganshin, Ayrat M; Schmidt, Thomas; Lv, Zuopeng; Liebetrau, Jan; Richnow, Hans Hermann; Kleinsteuber, Sabine; Nikolausz, Marcell

    2016-10-01

    The effects of hydraulic retention time (HRT) reduction at constant high organic loading rate on the activity of hydrogen-producing bacteria and methanogens were investigated in reactors digesting thin stillage. Stable isotope fingerprinting was additionally applied to assess methanogenic pathways. Based on hydA gene transcripts, Clostridiales was the most active hydrogen-producing order in continuous stirred tank reactor (CSTR), fixed-bed reactor (FBR) and anaerobic sequencing batch reactor (ASBR), but shorter HRT stimulated the activity of Spirochaetales. Further decreasing HRT diminished Spirochaetales activity in systems with biomass retention. Based on mcrA gene transcripts, Methanoculleus and Methanosarcina were the predominantly active in CSTR and ASBR, whereas Methanosaeta and Methanospirillum activity was more significant in stably performing FBR. Isotope values indicated the predominance of aceticlastic pathway in FBR. Interestingly, an increased activity of Methanosaeta was observed during shortening HRT in CSTR and ASBR despite high organic acids concentrations, what was supported by stable isotope data. Copyright © 2016 Elsevier Ltd. All rights reserved.

  13. 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

  14. Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

    International Nuclear Information System (INIS)

    Demachi, K.; Yoshida, Y.; Miya, K.

    1994-01-01

    For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

  15. 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

  16. 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)

    1999-12-31

    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)

  17. 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)

  18. 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

  19. Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

    1976-05-01

    For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

  20. The role of nuclear reactors in space exploration and development

    Energy Technology Data Exchange (ETDEWEB)

    Lipinski, R.J.

    2000-07-01

    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 {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 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

  1. 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.)

  2. Self-sustainability of a research reactor facility with neutron activation analysis

    International Nuclear Information System (INIS)

    Chilian, C.; Kennedy, G.

    2010-01-01

    Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

  3. 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

  4. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  5. Horseshoes in a Chaotic System with Only One Stable Equilibrium

    Science.gov (United States)

    Huan, Songmei; Li, Qingdu; Yang, Xiao-Song

    To confirm the numerically demonstrated chaotic behavior in a chaotic system with only one stable equilibrium reported by Wang and Chen, we resort to Poincaré map technique and present a rigorous computer-assisted verification of horseshoe chaos by virtue of topological horseshoes theory.

  6. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  7. Removal of uranium and salt from the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

    1998-01-01

    In 1994, migration of 233 U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage

  8. Removal of uranium and salt from the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

    1998-06-01

    In 1994, migration of {sup 233}U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage.

  9. Mixing of Al into uranium silicides reactor fuels

    International Nuclear Information System (INIS)

    Ding, F.R.; Birtcher, R.C.; Kestel, B.J.; Baldo, P.M.

    1996-11-01

    SEM observations have shown that irradiation induced interaction of the aluminum cladding with uranium silicide reactor fuels strongly affects both fission gas and fuel swelling behaviors during fuel burn-up. The authors have used ion beam mixing, by 1.5 MeV Kr, to study this phenomena. RBS and the 27 Al(p, γ) 28 Si resonance nuclear reaction were used to measure radiation induced mixing of Al into U 3 Si and U 3 Si 2 after irradiation at 300 C. Initially U mixes into the Al layer and Al mixes into the U 3 Si. At a low dose, the Al layer is converted into UAl 4 type compound while near the interface the phase U(Al .93 Si .07 ) 3 grows. Under irradiation, Al diffuses out of the UAl 4 surface layer, and the lower density ternary, which is stable under irradiation, is the final product. Al mixing into U 3 Si 2 is slower than in U 3 Si, but after high dose irradiation the Al concentration extends much farther into the bulk. In both systems Al mixing and diffusion is controlled by phase formation and growth. The Al mixing rates into the two alloys are similar to that of Al into pure uranium where similar aluminide phases are formed

  10. Research works at the Physics Institute nuclear reactor for the nuclear power engineering

    International Nuclear Information System (INIS)

    Gavars, V.V.; Kalnin'sh, D.O.; Lapenas, A.A.; Tomsons, E.Ya.; Ulmanis, U.A.

    1985-01-01

    Methods for neutron spectra determination in the nuclear reactor core and vessel have been developed. On their base the neutron spectra at the Novo-Voronezh and kola NPPs have been measured. Such measurements are necessary for the determination of the nuclear fuel reprocessing coefficients, for the evaluation of the construction radiation-damage stability and the NPP economical efficiency on the whole. A new type of the reactor regulator - a liquid metal one - has been created. Such regulators are promising in respect of their use at the NPPs. The base for studying new radiation-damage-stable insulators has been founded. The materials obtained are now applied to designing the reactors of the second (fast) and the third (thermonuclear H) generations. There have developed and by a long-time exploitation checked a hot loop, used for materials irradiation. the nuclear reactor in Salaspils provides training of students being the new brain-power for the nuclear power engineering

  11. 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

  12. Molten salt reactors and possible scenarios for future nuclear power deployment

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

    2004-01-01

    An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  14. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  15. Dynamic simulation of a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant

  16. Direct transition from a stable equilibrium to quasiperiodicity in non-smooth systems

    DEFF Research Database (Denmark)

    Zhusubaliyev, Z.T.; Mosekilde, Erik

    2008-01-01

    The purpose of this Letter is to show how a border-collision bifurcation in a piecewise-smooth dynamical system can produce a direct transition from a stable equilibrium point to a two-dimensional invariant torus. Considering a system of nonautonomous differential equations describing the behavior...... of a power electronic DC/DC converter, we first determine the chart of dynamical modes and show that there is a region of parameter space in which the system has a single stable equilibrium point. Under variation of the parameters, this equilibrium may collide with a discontinuity boundary between two smooth...... regions in phase space. When this happens, one can observe a number of different bifurcation scenarios. One scenario is the continuous transformation of the stable equilibrium into a stable period-1 cycle. Another is the transformation of the stable equilibrium into an unstable period-1 cycle with complex...

  17. Application of a PID controller based on fuzzy logic to reduce variations in the control parameters in PWR reactors

    International Nuclear Information System (INIS)

    Vasconcelos, Wagner Eustaquio de; Lira, Carlos Alberto Brayner de Oliveira; Brito, Thiago Souza Pereira de; Afonso, Antonio Claudio Marques; Cruz Filho, Antonio Jose da; Marques, Jose Antonio; Teixeira, Marcello Goulart

    2013-01-01

    Nuclear reactors are in nature nonlinear systems and their parameters vary with time as a function of power level. These characteristics must be considered if large power variations occur in power plant operational regimes, such as in load-following conditions. A PWR reactor has a component called pressurizer, whose function is to supply the necessary high pressure for its operation and to contain pressure variations in the primary cooling system. The use of control systems capable of reducing fast variations of the operation variables and to maintain the stability of this system is of fundamental importance. The best-known controllers used in industrial control processes are proportional-integral-derivative (PID) controllers due to their simple structure and robust performance in a wide range of operating conditions. However, designing a fuzzy controller is seen to be a much less difficult task. Once a Fuzzy Logic controller is designed for a particular set of parameters of the nonlinear element, it yields satisfactory performance for a range of these parameters. The objective of this work is to develop fuzzy proportional-integral-derivative (fuzzy-PID) control strategies to control the level of water in the reactor. In the study of the pressurizer, several computer codes are used to simulate its dynamic behavior. At the fuzzy-PID control strategy, the fuzzy logic controller is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region. Thus the fuzzy logic controller tunes the gain of PID controller to adapt the model with changes in the water level of reactor. The simulation results showed a favorable performance with the use to fuzzy-PID controllers. (author)

  18. Modelling of temperature distribution and temperature pulsations in elements of fast breeder reactor

    International Nuclear Information System (INIS)

    Sorokin, A.P.; Bogoslovskaia, G.P.; Ushakov, P.A.; Zhukov, A.V.; Ivanov, Eu.F.; Matjukhin, N.M.

    2004-01-01

    From thermophysical point of view, integrated configuration of liquid metal cooled reactor has some limitations. Large volume of mixing chamber causes a complex behavior of thermal hydraulic characteristics in such facilities. Also, this volume is responsible for large-scale eddies in the coolant, existence of stagnant areas and flow stratification, occurrence of temperature non-uniformity and pulsation of coolant and structure temperatures. Temperature non-uniformities and temperature pulsations depend heavily even on small variations in reactor core design. The paper presents some results on modeling of thermal hydraulic processes occurring in liquid metal cooled reactor. The behavior of following parameters are discussed: temperature non-uniformities at the core output and related temperature pulsations; temperature pulsations due to mixing of sodium jets at different temperatures; temperature pulsations arising if a part of loop (circuit) is shut off; temperature non-uniformities and pulsation at the core output and related temperature pulsation; temperature pulsations due to mixing of sodium jets at different temperatures; temperature pulsations arising if a part of loop (circuit) is shut off; temperature non-uniformities and pulsation of temperature during transients and during transition to natural convection cooling. Also, the issue of modeling of temperature behavior in compact arrangement of fast reactor fuel pins using water as modeling liquid is considered in the paper. One more discussion is concerned with experimental method of modeling of liquid metal mixing with the use of air. The method is based on freon tracer technique. The results of simulation of the thermal hydraulic processes mentioned above have been analyzed, that will allow the main lines of the study to be determined and conclusion to be drawn regarding the temperature behavior in fast reactor units. (author)

  19. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon

    2006-02-01

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation

  20. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.