Fast reactor safety and related physics. Volume I. Invited papers; panels; summary

Energy Technology Data Exchange (ETDEWEB)

1976-01-01

Separate abstracts were prepared for each of the twenty invited papers included. The papers covered sessions on licensing aspects of safety design bases, safety of demonstration plants, safety aspects of large commercial fast breeders, and safety test facilities.

Behaviour of molten reactor fuels under accident conditions

International Nuclear Information System (INIS)

Xavier Swamikannu, A.; Mathews, C.K.

1980-01-01

The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

Air conditioning device for reactor buildings

International Nuclear Information System (INIS)

Kikuchi, Shiro.

1982-01-01

Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

Cable condition monitoring in a pressurized water reactor environment

International Nuclear Information System (INIS)

Al-Hussaini, T.J.

1988-01-01

Oconee Nuclear Station is the first nuclear plant designed, engineered and constructed by Duke Power Company. Even though the accelerated aging method was available to determine the life expectancy of the cable used in the reactor building, no natural aging data was available at that time. In order to be able to verify the condition of the reactor building cable over the life of the plant, an on-going cable monitoring plan was instituted. Various types of cable were selected to be monitored, and they were installed in cable life evaluation circuits in the reactor building. At five year intervals over the life of the plant, cable samples would be removed from these cable life evaluation circuits and tested to determine the effects of the reactor building environment on the integrity of the cable. A review of the cable life evaluation circuits and the results of the evaluation program to date is presented

Research on reactor physics data

International Nuclear Information System (INIS)

1961-01-01

In the early years of nuclear reactor research, each national program tended to develop its own reactor physics information. The Government of Norway proposed to the Agency the undertaking of a joint program in reactor physics utilizing the facilities and staff of its zero power reactor NORA then under construction. Following the approval by the Board of Governors in February, the Agency invited Member States to submit the names and qualifications of scientists they wished to suggest for the project. All the results and information gained through the program, which is expected to last about three years, will be placed at the disposal of the Agency's Member States

Power conditioning for space nuclear reactor systems

Science.gov (United States)

Berman, Baruch

1987-01-01

This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

International Nuclear Information System (INIS)

Es-Saheb, M.H.H.

1990-01-01

The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

The IAEA's international project on innovative nuclear reactors and fuel cycles (INPRO)

International Nuclear Information System (INIS)

Kuptiz, Juergen; )

2002-01-01

This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited all interested Member states to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology and invited Member states to consider to contribute to a task force on innovative nuclear reactors and fuel cycle

Reactor internals design/analysis for normal, upset, and faulted conditions

International Nuclear Information System (INIS)

Burke, F.R.

1977-06-01

The analytical procedures used by Babcock and Wilcox to demonstrate the structural integrity of the 205-FA reactor internals are described. Analytical results are presented and compared to ASME Code allowable limits for Normal, Upset, and Faulted conditions. The particular faulted condition considered is a simultaneous loss-of-coolant accident and safe shutdown earthquake. The operating basis earthquake is addressed as an Upset condition

Wireless condition monitoring for the RA-6 research reactor

International Nuclear Information System (INIS)

Garcia Peyrano, O.; Calzeta, O.; Rico, N.; Damiani, H.; Coutsiers, E.

1999-01-01

The vibration laboratory at C.A.B. has a great experience with the analysis and diagnostic of symptoms of failures in the rotating equipment of the R-6 research reactor and in our longest NPP (CANDU 600 Mw), located in Embalse town, Cordoba City, Argentina. Objective: The standard condition monitoring instrumentation system were designed for large equipment operating under different environmental conditions and sensitivities. The signal processing is not flexible and the diagnostic is an expensive method for the small poll type research reactors. This papers describes the research and development which are related whit the new concept, cheaper and flexible condition monitoring instrumentation system. Implementing a vibration analysis measurements technique with a sensor inside (in the pool) of the nuclear reactor RA-6, and mainly based on fft signal processing, an extensive program for vibration source identification was done. Different nuclear power conditions were monitored as full power and in zero power, also. This zero power shows the best acoustical environmental, because the cooling pumps are stop, and the core is cooling by natural convection. Two sensors were mainly used as the detector's subsystem. One of these detectors was an accelerometer attached to the top of the fine control rod and the other one was a water resistant omnidirectional microphone which was located underwater at different distances from the nuclear core. All the signal measurement by this two sensors were recorded and then was processed. Both signal was acquired at the same time for correlation analysis purposes. The analysis was composed by a 'Spectral Dynamics SD380' connected to a P.C. with dedicated post processing software. On the other hand, some calibration and sensitivity comparison was done using an SKFCM40, dual channel data collector and analyzer. (author)

Pattern recognition application for surveillance of abnormal conditions in a nuclear reactor

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Dzwinel, W.

1990-01-01

The system to monitor abnormal conditions in a nuclear reactor, based on the noise analysis of the reactor basic parameters such as power, temperature and coolant flow rate, has been developed. The pattern recognition techniques such as clustering, cluster analysis, feature selection and clusters visualization methods form the basis of the software. Apart from non-hierarchical clustering procedures applied earlier, the hierarchical one is recommended. The system application for IBR-2 Dubna reactor diagnostics is shown. 10 refs.; 6 figs

Conditioning for definitive storage of radioactive graphite bricks from reactor decommissioning

International Nuclear Information System (INIS)

Costes, J.R.; Koch, C.; Tassigny, C. de; Vidal, H.; Raymond, A.

1990-01-01

The decommissioning of gas-graphite reactors in the EC (e.g. French UNGGs, British Magnox reactors and AGRs, and reactors in Spain and in Italy) will produce large amounts of graphite bricks. This graphite cannot be accepted without particular conditioning by the existing shallow land disposal sites. The aim of the study is to examine the behaviour of graphite waste and to develop a conditioning technique which makes this waste acceptable for shallow land disposal sites. 18 kg of graphite core samples with an outside diameter of 74 mm were removed from the G2 gas-cooled reactor at Marcoule. Their radioactivity is highly dependent on the position of the graphite bricks inside the reactor. Measured results indicate an activity range of 100-400 MBq/kg with 90% Tritium, 5% 14 C, 3% 60 Co, 1.5% 63 Ni. Repeated porosity analyses showed that open porosity ranging from 0 to 100 μm exceeded 23 vol% in the graphite. Water penetration kinetics were investigated in unimpregnated graphite and resulted in impregnation by water of 50-90% of the open porosity. Preliminary lixiviation tests on the crude samples showed quick lixidegree of Cs (several per cent) and of 60 Co, and 133 Ba at a lesser degree. The proposed conditioning technique does not involve a simple coating but true impregnation by a tar-epoxy mixture. The bricks recovered intact from the core by robot services will be placed one by one inside a cylindrical metallic container. But this container may corrode and the bricks may become fragmented in the future, the normally porous graphite will be unaffected by leaching since it is proved that all pores larger than 0.1 μm will be filled with the tar-epoxy mixture. This is a true long-term waste packaging concept. The very simple technology required for industrial implementation is discussed

A study of passive safety conditions for fast reactor core

International Nuclear Information System (INIS)

Shimizu, Akinao

1991-01-01

A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

Behaviour of gas cooled reactor fuel under accident conditions

International Nuclear Information System (INIS)

1991-11-01

The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

Scram reliability under seismic conditions at the Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Roglans, J.; Wang, C.Y.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

International Nuclear Information System (INIS)

Hodge, S.A.

1990-01-01

Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

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

Science.gov (United States)

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

2012-05-01

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

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

Directory of Open Access Journals (Sweden)

M. Weclas

2013-01-01

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

Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

International Nuclear Information System (INIS)

Kim, Hak Jae; Park, Goon Cherl

1996-01-01

Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

International Nuclear Information System (INIS)

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

2014-01-01

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

Optimal conditions in direct dimethyl ether synthesis from syngas utilizing a dual-type fluidized bed reactor

International Nuclear Information System (INIS)

Yousefi, Ahmad; Eslamloueyan, Reza; Kazerooni, Nooshin Moradi

2017-01-01

Concerns over environmental pollution and ever-increasing energy demand have urged the global community to tap clean-burning fuels among which dimethyl ether is a promising candidate for contribution in the transportation sector. Direct dimethyl ether synthesis from syngas, in which methanol production and dehydration take place simultaneously, is arguably the preferred route for large scale production. In this study, direct dimethyl ether synthesis is proposed in an industrial dual-type fluidized bed reactor. This configuration involves two fluidized bed reactors operating in different conditions. In the first catalytic reactor (water-cooled reactor), the synthesis gas is partly converted to methanol after being preheated by the reaction heat in the second reactor (gas-cooled reactor). A two-phase generalized comprehensive reactor model, comprised of the flow in three different regimes is applied and a smooth transition between flow regimes is provided based on the probabilistic averaging approach. The optimal operating conditions are sought by employing differential evolution algorithm as a robust optimization strategy. The dimethyl ether mole fraction is considered as the objective function during the optimization. The results show considerable dimethyl ether enhancement by 16% and 14% compared to the conventional direct dimethyl ether synthesis reactor and dual-type fixed bed dimethyl ether reactor arrangements, respectively. - Highlights: • Dual-type catalytic fluidized bed reactors for dimethyl ether synthesis is studied. • A two-phase comprehensive model comprised of flow in three regimes is used. • Probabilistic averaging approach is applied for smooth transitions between regimes. • Differential evolution method is employed to determine optimal operating conditions. • Production capacity is remarkably enhanced compared to conventional reactor.

Effect of reactor finiteness on the boundary condition at the surface of a booster section

International Nuclear Information System (INIS)

Wassef, W.A.

1982-01-01

Effect of reactor finiteness on the boundary condition at the surface of an absorbing booster embedded in the reactor core is studied and formulated. The model used in these calculations depends on the Pl-Transport coupling technique. This method takes into consideration the rigorous neutron transport behavior inside the booster medium, while the Pl-approximation in the bulk of the scattering medium surrounding the booster which can be considered infinite in most practical applications. The neutron flux gradient parallel to the surface of the booster is considered. The geometrical configuration of the reactor core cross section is circular or rectangular. Finiteness of the reactor is introduced in the general formulation through its dimensions or buckling. Extensive numerical results are given to demonstrate the dependence of the boundary condition at the surface of the booster section on the reactor finiteness and the different physical parameters

Heat transfer study of a submerged reactor channel under boil-off condition

Energy Technology Data Exchange (ETDEWEB)

Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

2012-12-15

Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

Helminth Egg Removal Capacity of UASB Reactors under Subtropical Conditions

Directory of Open Access Journals (Sweden)

Rosa-Elena Yaya-Beas

2015-05-01

Full Text Available This research was conducted to study the anaerobic sludge filtration capacity regarding helminth egg removal in upflow anaerobic sludge blanket (UASB reactors. Two 25 L lab-scale UASB reactors were operated at an ambient temperature which varied between 17.1 and 28.6 °C. Ascaris suum egg was selected as the model egg considering its similarity in terms of size and morphology to Ascaris lumbricoides, a human pathogen. Ascaris suum eggs were obtained from female parasites of infected pigs. The anaerobic sludge filtration capacity was performed applying upflow velocities between 0.09 and 0.68 m·h−1. Three sludge bed heights in the range of 0.30–0.40 m, 0.50–0.60 m and 0.60–0.70 m were applied. These sludge bed heights corresponded to 19%–25%, 31%–38% and 38%–44% of the total reactor height, respectively. Under the mentioned conditions, the average helminth egg removal efficiency was reciprocally correlated to the imposed upflow velocity. The studied lab-scale reactors reported an average helminth egg removal between 34%–100%, 30%–91% and 34%–56%, when the sludge bed in the UASB reactor was 19%–25%, 31%–38% and 38%–44% of the total reactor height, respectively. The decreased filtration capacity at increasing sludge bed heights might be likely related to biogas production and channeling formation. The average helminth egg removal efficiency in the control experiments performed without any sludge bed, by plain sedimentation, varied between 44% and 66%.

Electrochemical noise measurements under pressurized water reactor conditions

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2000-01-01

Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

Energy Technology Data Exchange (ETDEWEB)

Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

International Nuclear Information System (INIS)

Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu

2016-01-01

To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day -1 of air, 0.004%·day -1 of noble gas and 3.7×10 -5 %·day -1 of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m 3 ·hr -1 , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr -1 under the condition of 20 m·sec -1 wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor

Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited).

Science.gov (United States)

Smith, Roger J

2008-10-01

A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B(pol) diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T(e), n(e), and B(parallel) along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n(e)B(parallel) product and higher n(e) and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited)

International Nuclear Information System (INIS)

Smith, Roger J.

2008-01-01

A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B pol diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T e , n e , and B || along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n e B || product and higher n e and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

International Nuclear Information System (INIS)

Meng Qinghu; Meng Qingfeng; Feng Wuwei

2012-01-01

Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

Science.gov (United States)

Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

1988-01-01

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

2009-06-15

ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

Effect of redox conditions on pharmaceutical loss during biological wastewater treatment using sequencing batch reactors

Energy Technology Data Exchange (ETDEWEB)

Stadler, Lauren B., E-mail: lstadler@umich.edu [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States); Su, Lijuan, E-mail: lijuansu@buffalo.edu [Department of Chemistry, University at Buffalo, State University of New York, Buffalo, NY 14260 (United States); Moline, Christopher J., E-mail: christopher.moline@hdrinc.com [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States); Ernstoff, Alexi S., E-mail: alexer@dtu.dk [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States); Aga, Diana S., E-mail: dianaaga@buffalo.edu [Department of Chemistry, University at Buffalo, State University of New York, Buffalo, NY 14260 (United States); Love, Nancy G., E-mail: nglove@umich.edu [Department of Civil and Environmental Engineering, University of Michigan, 1351 Beal Avenue, EWRE, Ann Arbor, MI 48109 (United States)

2015-01-23

Highlights: • Pharmaceutical fate was studied in SBRs operated at different redox conditions. • Stable carbon oxidation and nitrification occurred under microaerobic conditions. • Losses of atenolol and trimethoprim were highest under fully aerobic conditions. • Loss of sulfamethoxazole was highest under microaerobic conditions. • Deconjugation occurred during treatment to form sulfamethoxazole and desvenlafaxine. - Abstract: We lack a clear understanding of how wastewater treatment plant (WWTP) process parameters, such as redox environment, impact pharmaceutical fate. WWTPs increasingly install more advanced aeration control systems to save energy and achieve better nutrient removal performance. The impact of redox condition, and specifically the use of microaerobic (low dissolved oxygen) treatment, is poorly understood. In this study, the fate of a mixture of pharmaceuticals and several of their transformation products present in the primary effluent of a local WWTP was assessed in sequencing batch reactors operated under different redox conditions: fully aerobic, anoxic/aerobic, and microaerobic (DO concentration ≈0.3 mg/L). Among the pharmaceuticals that were tracked during this study (atenolol, trimethoprim, sulfamethoxazole, desvenlafaxine, venlafaxine, and phenytoin), overall loss varied between them and between redox environments. Losses of atenolol and trimethoprim were highest in the aerobic reactor; sulfamethoxazole loss was highest in the microaerobic reactors; and phenytoin was recalcitrant in all reactors. Transformation products of sulfamethoxazole and desvenlafaxine resulted in the reformation of their parent compounds during treatment. The results suggest that transformation products must be accounted for when assessing removal efficiencies and that redox environment influences the degree of pharmaceutical loss.

Effect of redox conditions on pharmaceutical loss during biological wastewater treatment using sequencing batch reactors

International Nuclear Information System (INIS)

Stadler, Lauren B.; Su, Lijuan; Moline, Christopher J.; Ernstoff, Alexi S.; Aga, Diana S.; Love, Nancy G.

2015-01-01

Highlights: • Pharmaceutical fate was studied in SBRs operated at different redox conditions. • Stable carbon oxidation and nitrification occurred under microaerobic conditions. • Losses of atenolol and trimethoprim were highest under fully aerobic conditions. • Loss of sulfamethoxazole was highest under microaerobic conditions. • Deconjugation occurred during treatment to form sulfamethoxazole and desvenlafaxine. - Abstract: We lack a clear understanding of how wastewater treatment plant (WWTP) process parameters, such as redox environment, impact pharmaceutical fate. WWTPs increasingly install more advanced aeration control systems to save energy and achieve better nutrient removal performance. The impact of redox condition, and specifically the use of microaerobic (low dissolved oxygen) treatment, is poorly understood. In this study, the fate of a mixture of pharmaceuticals and several of their transformation products present in the primary effluent of a local WWTP was assessed in sequencing batch reactors operated under different redox conditions: fully aerobic, anoxic/aerobic, and microaerobic (DO concentration ≈0.3 mg/L). Among the pharmaceuticals that were tracked during this study (atenolol, trimethoprim, sulfamethoxazole, desvenlafaxine, venlafaxine, and phenytoin), overall loss varied between them and between redox environments. Losses of atenolol and trimethoprim were highest in the aerobic reactor; sulfamethoxazole loss was highest in the microaerobic reactors; and phenytoin was recalcitrant in all reactors. Transformation products of sulfamethoxazole and desvenlafaxine resulted in the reformation of their parent compounds during treatment. The results suggest that transformation products must be accounted for when assessing removal efficiencies and that redox environment influences the degree of pharmaceutical loss

Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

2013-10-15

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

Directory of Open Access Journals (Sweden)

VIKAS JAIN

2013-10-01

Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

Invitation to View Method with Advantages and Disadvantages

Directory of Open Access Journals (Sweden)

Selçuk BUYRUKOĞLU

2018-01-01

Full Text Available Taxpayers and tax administrators sometimes may experience tax dispute. These disputes may arise from the taxpayers as well as from the tax administrators. In resolving tax disputes, tax administrators and tax payers are the first, usually administratively, solution. Because while the tax revenues of the state are possible entered into the state treasury, the taxpayers will benefit from the advantages of administrative solutions. The state and taxpayer herein will be the winner. Invitation to view is one of the administrative remedies for tax disputes. With this practice taxpayers who have caused tax evasion penalty are invited by tax administrators to make a declaration on the condition that the tax examination of the tax disputes is not started and the dispatch process has not been done to the discretion comission. In this study, advantages and disadvantages will be addressed after the definition, legal support, purpose, scope and mechanism of Invitation to view.

Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emissions in an Oxidation Flow Reactor

Science.gov (United States)

Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

International Nuclear Information System (INIS)

Philpott, E.F.; Pounder, F.; Willby, C.R.

1978-01-01

The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

Operational limits and conditions and operating procedures for research reactors. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This publication provides practical guidance on all important aspects of developing, formulating and presenting the operational limits and conditions as well as the operating procedures for research reactors. It covers the concept of operational limits and conditions, their content, and the responsibilities of the operating organization with respect to their establishment, modification, documentation and compliance. The guidance also covers the training of operating personnel on performing periodic testing, established by the operational limits and conditions, and operating procedures

Upflow anaerobic sludge reactors for the treatment of combined industrial effluent in subtropical conditions: a comparison between UASB and UASF reactors

International Nuclear Information System (INIS)

Yasar, A.; Ahmad, N.; Chaudhry, M.N.; Sarwar, M.; Masood, T.; Yaqub, A.

2005-01-01

The performance of anaerobic biological process is heavily process conditions dependent. In this study, an attempt has been made to investigate the influence of process conditions like temperature, sludge age and hydraulic retention time (HRT) on the efficiency of an upflow anaerobic sludge blanket (UASB) reactor and upflow anaerobic sludge filter (UASF) to treat combined industrial wastewater. Reactors were operated at easing ambient temperatures (38, 30, 20 and 14 deg. C) and correspondingly increasing sludge ages (60, 90, 120 and 150 days). At temperature 38 deg. C and sludge age of 60 days, UASF showed better performance than VASE reactor. This mainly due to the enhanced filtration through well-graded sand filter and fairly good biological activity in UASF. At this stage, lack of sludge granulation in VASE reactor resulted in poor biological activity; hence, relatively poor performance. At temperatures 30 and 20 deg. C with sludge ages of 90 and 120 days, respectively, UASB gave better results than UASF. The reason was rapid biological degradation due to proper sludge granulation and favorable temperature. At temperature 14 deg. C, a substantial decrease in the efficiency of UASB reactor as compared to the UASF was evident. Drop in efficiency was because of inhabitation of methanogenic bacteria and liquidation of sludge granules. These factors mounted to a decrease in biological activity, stoppage as production and an increase in total suspended solids (TSS) in the effluent. The influence of hydraulic retention time (ranging between 3-12 hours at an increment of 3 hours) on the removal efficiency of both UASB and UASF was not significant. At favorable temperature (20 to 30 deg. C) and sludge age (90 to 120 days) UASB reactor appeared to be more efficient than UASF.(author)

Corrosion fatigue initiation and short crack growth behaviour of austenitic stainless steels under light water reactor conditions

International Nuclear Information System (INIS)

Seifert, H.P.; Ritter, S.; Leber, H.J.

2012-01-01

Highlights: ► Corrosion fatigue in austenitic stainless steels under light water reactor conditions. ► Identification of major parameters of influence on initiation and short crack growth. ► Critical system conditions for environmental reduction of fatigue initiation life. ► Comparison with the environmental factor (F env ) approach. - Abstract: The corrosion fatigue initiation and short crack growth behaviour of different wrought low-carbon and stabilised austenitic stainless steels was characterised under simulated boiling water reactor and pressurised water reactor primary water conditions by cyclic fatigue tests with sharply notched fracture mechanics specimens. The special emphasis was placed to the behaviour at low corrosion potentials and, in particular, to hydrogen water chemistry conditions. The major parameter effects and critical conjoint threshold conditions, which result in relevant environmental reduction and acceleration of fatigue initiation life and subsequent short crack growth, respectively, are discussed and summarised. The observed corrosion fatigue behaviour is compared with the fatigue evaluation procedures in codes and regulatory guidelines.

Possibilities of crack mouth opening displacement (CMOD) measurement under boiling and pressurized water reactor conditions

International Nuclear Information System (INIS)

Ehling, W.

1984-01-01

Fracture mechanics investigations carried out so far in laboratory conditions cover only part of the material stresses, as effects which occur in nuclear powerstations, in particular, such as corrosion and radioactive radiation are largely left out of account. Therefore experiments including these effects were recently carried out in autoclaves, test rigs simulating reactors (HRD experimental plant) and in experimental reactors. An important parameter of experimental fracture mechanics is the measurement of crack opening displacement (COD). The crack opening is measured with socalled clip gauges (transmitters based on strain gauges, which convert mechanical deformation of springs into electrical signals) on standard samples in the laboratory. It was therefore sensible to use these high temperature strain gauges (HTD) for the development of a measuring system for travel for pressurized water and boiling water reactor conditions. (orig.) [de

A matrix formalism to solve interface condition equations in a reactor system

Energy Technology Data Exchange (ETDEWEB)

Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1970-05-15

When a nuclear reactor or a reactor lattice cell is treated by an approximate procedure to solve the neutron transport equation, as the last computational step often appears a problem of solving systems of algebraic equations stating the interface and boundary conditions for the neutron flux moments. These systems have usually the coefficient matrices of the block-bi diagonal type, containing thus a large number of zero elements. In the present report it is shown how such a system can be solved efficiently accounting for all the zero elements both in the coefficient matrix and in the free term vector. The procedure is presented here for the case of multigroup P{sub 3} calculation of neutron flux distribution in a cylindrical reactor lattice cell. Compared with the standard gaussian elimination method, this procedure is more advantageous both in respect to the number of operations needed to solve a given problem and in respect to the computer memory storage requirements. A similar formalism can also be applied for other approximate methods, for instance for multigroup diffusion treatment of a multi zone reactor. (author)

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

Directory of Open Access Journals (Sweden)

Muhammad Darwis Isnaini

2016-03-01

Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-10-01

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

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

Tools and techniques for ageing predictions in nuclear reactors through condition monitoring

International Nuclear Information System (INIS)

Verma, R.M.P.

1994-01-01

To operate the nuclear reactors beyond their design predicted life is gaining importance because of huge replacement and decommissioning costs. But experience shows that nuclear plant safety and reliability may decline in the later years of plant life due to ageing degradation. Ageing of nuclear plant components, structures and systems, if unmitigated reduces their safety margins provided in the design and thus increases risks to public health and safety. These safety margins must be monitored throughout plant service life including any extended life. Condition monitoring of nuclear reactor components/equipment and systems can be done to study the effect of ageing, status of safety margins and effect of corrective and mitigating actions taken. The tools and techniques of condition monitoring are also important in failure trending, predictive maintenance, evaluation of scheduled maintenance, in mitigation of ageing, life extension and reliability studies. (author). 1 fig., 1 annexure

Design requirements, operation and maintenance of gas-cooled reactors

International Nuclear Information System (INIS)

1989-06-01

At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

Criticality safety studies involved in actions to improve conditions for storing 'RA' research reactor spent fuel

International Nuclear Information System (INIS)

Matausek, M.; Marinkovic, N.

1998-01-01

A project has recently been initiated by the VINCA Institute of Nuclear Sciences to improve conditions in the spent fuel storage pool at the 6.5 MW research reactor RA, as well as to consider transferring this spent fuel into a new dry storage facility built for the purpose. Since quantity and contents of fissile material in the spent fuel storage at the RA reactor are such that possibility of criticality accident can not be a priori excluded, according to standards and regulations for handling fissile material outside a reactor, before any action is undertaken subcriticality should be proven under normal, as well as under credible abnormal conditions. To perform this task, comprehensive nuclear criticality safety studies had to be performed. (author)

Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

International Nuclear Information System (INIS)

Mitari, O.; Hirose, A.; Skarsgard, H.M.

1989-01-01

In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

International Nuclear Information System (INIS)

Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

2015-01-01

Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

Summary of the fifth international conference on reactor shielding

International Nuclear Information System (INIS)

Roussin, R.W.; Abbott, L.S.; Bartine, D.E.

1977-01-01

The Fifth International Conference on Reactor Shielding was held April 18-23, 1977 in Knoxville, Tennessee. The meeting was the largest in the series and attracted participants from 34 countries. The 10 invited papers and 10 of the contributed papers, selected as being representative of the Conference by the Technical Program Committee, are published in this issue of ATOMKERNENERGIE. This collection of papers demonstrates that the field of nuclear reactor shielding has developed into a mature discipline while retaining a definite vitality. (orig.) [de

Invitational Education: Theory, Research and Practice

Science.gov (United States)

Haigh, Martin

2011-01-01

Invitational Theory argues that learning is enhanced when learners are positively encouraged or "invited" into the educational experience. Arising from perceptual and self-concept theory, Invitational Pedagogy is constructed on four principles: respect for people, trust, optimism and intentionality, and upon five pillars: people, places, policies,…

Effective utilization and management of research reactors

Energy Technology Data Exchange (ETDEWEB)

Muranaka, R [International Atomic Energy Agency, Vienna (Austria). Div. of Research and Isotopes

1984-06-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors.

Effective utilization and management of research reactors

International Nuclear Information System (INIS)

Muranaka, R.

1984-01-01

The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

International Nuclear Information System (INIS)

Roberto, Thiago D.; Alvim, Antonio C.M.

2017-01-01

The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

Energy Technology Data Exchange (ETDEWEB)

Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

Temperature conditions of foundation plates under nuclear power plant reactor compartments

International Nuclear Information System (INIS)

Ehsaulov, S.L.

1990-01-01

Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

Bioremediation of endosulfan contaminated soil and water-Optimization of operating conditions in laboratory scale reactors

International Nuclear Information System (INIS)

Kumar, Mathava; Philip, Ligy

2006-01-01

A mixed bacterial culture consisted of Staphylococcus sp., Bacillus circulans-I and -II has been enriched from contaminated soil collected from the vicinity of an endosulfan processing industry. The degradation of endosulfan by mixed bacterial culture was studied in aerobic and facultative anaerobic conditions via batch experiments with an initial endosulfan concentration of 50 mg/L. After 3 weeks of incubation, mixed bacterial culture was able to degrade 71.58 ± 0.2% and 75.88 ± 0.2% of endosulfan in aerobic and facultative anaerobic conditions, respectively. The addition of external carbon (dextrose) increased the endosulfan degradation in both the conditions. The optimal dextrose concentration and inoculum size was estimated as 1 g/L and 75 mg/L, respectively. The pH of the system has significant effect on endosulfan degradation. The degradation of alpha endosulfan was more compared to beta endosulfan in all the experiments. Endosulfan biodegradation in soil was evaluated by miniature and bench scale soil reactors. The soils used for the biodegradation experiments were identified as clayey soil (CL, lean clay with sand), red soil (GM, silty gravel with sand), sandy soil (SM, silty sand with gravel) and composted soil (PT, peat) as per ASTM (American society for testing and materials) standards. Endosulfan degradation efficiency in miniature soil reactors were in the order of sandy soil followed by red soil, composted soil and clayey soil in both aerobic and anaerobic conditions. In bench scale soil reactors, endosulfan degradation was observed more in the bottom layers. After 4 weeks, maximum endosulfan degradation efficiency of 95.48 ± 0.17% was observed in red soil reactor where as in composted soil-I (moisture 38 ± 1%) and composted soil-II (moisture 45 ± 1%) it was 96.03 ± 0.23% and 94.84 ± 0.19%, respectively. The high moisture content in compost soil reactor-II increased the endosulfan concentration in the leachate. Known intermediate metabolites of

Model experiments on simulation of the WWER water-chemical conditions at loop facilities of the MIR reactor

International Nuclear Information System (INIS)

Benderskaya, O.S.; Zotov, E.A.; Kuprienko, V.A.; Ovchinnikov, V.A.

1999-01-01

The experiments on simulation of the WWER type reactors water-chemical conditions have been started at the State Scientific Center RIAR. These experiments are being conducted at the multi-loop research MIR reactor at the PVK-2 loop facility. The dosage stand was created. It allows introduction of boric acid, potassium and lithium hydroxides, ammonia solutions and gaseous hydrogen. Corrosion tests of the Russian E-635 and E-110 alloys are being conducted at the PVK-2 loop under the WWER water-chemical conditions. If necessary, fuel elements are periodically extracted from the reactor to perform visual examination, to measure their length, diameter, to remove the deposits from the claddings, to measure the burnup and to distribute the fission products over the fuel element by gamma-spectrometry. The chemical analytical 'on line' equipment produced by the ORBISPHERE Laboratory (Switzerland) will be commissioned in the nearest future to measure concentration of the dissolved hydrogen and oxygen as well as pH and specific conductivity. The objective of the report is to familiarize the participants of the IAEA Technical Committee with the capabilities of performing the model water-chemical experiments under the MIR reactor loop facility conditions. (author)

Transactions of the Twentieth Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-10-01

This report contains summaries of papers on reactor safety research to be presented at the 20th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 21--23, 1992. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included

Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2

International Nuclear Information System (INIS)

Nikolic, M.

1963-01-01

It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events [sr

Luncheon address: Development of the CANDU reactor

International Nuclear Information System (INIS)

Bain, A.S.

1997-01-01

The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

International Nuclear Information System (INIS)

Moreira, Joao Manoel Losada; Frajndlich, Roberto

1995-01-01

This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-12-01

At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

Failure strains and proposed limit strains for an reactor pressure vessel under severe accident conditions

International Nuclear Information System (INIS)

Krieg, R.

2005-01-01

The local failure strains of essential design elements of a reactor vessel are investigated. The size influence of the structure is of special interest. Typical severe accident conditions including elevated temperatures and dynamic loads are considered. The main part of work consists of test families with specimens under uniaxial and biaxial load. Within one test family the specimen geometry and the load conditions are similar, but the size is varied up to reactor dimensions. Special attention is given to geometries with a hole or a notch causing non-uniform stress and strain distributions typical for the reactor vessel. A key problem is to determine the local failure strain. Here suitable methods had to be developed including the so-called 'vanishing gap method', and the 'forging die method'. They are based on post-test geometrical measurements of the fracture surfaces and reconstructions of the related strain fields using finite element models. The results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Local failure strains could be determined. The values are rather high and size dependent. Statistical evaluation allow the proposal of limit strains which are also size dependent. If these limit strains are not exceeded, the structures will not fracture

Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

International Nuclear Information System (INIS)

Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

2004-01-01

One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

Energy Technology Data Exchange (ETDEWEB)

Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

2004-07-01

One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

Eastern Sources of Invitational Education.

Science.gov (United States)

Ryback, David

1993-01-01

Presents historical perspective suggesting that invitational theory shares many beliefs with ancient Eastern philosophies. Submits that teachers and other educators who embrace the invitational perspective may benefit from an understanding of Eastern principles. Briefly describes Buddhism, Confucianism, Taoism, Zen Buddhism, and their relevance to…

11 CFR 100.137 - Invitations, food, and beverages.

Science.gov (United States)

2010-01-01

... 11 Federal Elections 1 2010-01-01 2010-01-01 false Invitations, food, and beverages. 100.137...) Exceptions to Expenditures § 100.137 Invitations, food, and beverages. The cost of invitations, food, and... invitations, food and beverages provided by the individual on behalf of the candidate does not exceed $1,000...

11 CFR 100.77 - Invitations, food, and beverages.

Science.gov (United States)

2010-01-01

... 11 Federal Elections 1 2010-01-01 2010-01-01 false Invitations, food, and beverages. 100.77...) Exceptions to Contributions § 100.77 Invitations, food, and beverages. The cost of invitations, food and... invitations, food and beverages provided by the individual on behalf of the candidate does not exceed $1,000...

Power distributions in a nuclear reactor core under various operating conditions and their safety implications

International Nuclear Information System (INIS)

Yang, C.Y.; Cho, N.Z.

1988-01-01

A nonlinear reactor model has been developed considering typical feedback effects such as the moderator and fuel temperatures, xenon, and soluble boron. The resulting equation belongs to a class of nonlinear boundary value problems, and it was numerically shown using the arc-length continuation method that the nonlinear feedback equation may have multiple steady-state solutions due to existence of branch points for a range of parameters and feedback coefficients which are determined by various design and operating conditions. Solution diagrams of three distinct types were obtained depending on the values of feedback constants and stability was characterized for each solution. The starting point of instability (first limit point) in the solution diagram decreases as the temperature coefficients become large or the boron concentration increases, which means decrease of stability. When solution diagrams which are divided into stable and unstable regions are obtained for a wide range of feedback coefficients, we can choose the range of the reactor design and operating conditions such that the reactor does not encounter unstable situations

Effect of redox conditions on pharmaceutical loss during biological wastewater treatment using sequencing batch reactors

DEFF Research Database (Denmark)

Stadler, Lauren B.; Su, Lijuan; Moline, Christopher J.

2015-01-01

We lack a clear understanding of how wastewater treatment plant (WWTP) process parameters, such as redox environment, impact pharmaceutical fate. WWTPs increasingly install more advanced aeration control systems to save energy and achieve better nutrient removal performance. The impact of redox...... under different redox conditions: fully aerobic, anoxic/aerobic, and microaerobic (DO concentration ≈0.3 mg/L). Among the pharmaceuticals that were tracked during this study (atenolol, trimethoprim, sulfamethoxazole, desvenlafaxine, venlafaxine, and phenytoin), overall loss varied between them...... and between redox environments. Losses of atenolol and trimethoprim were highest in the aerobic reactor; sulfamethoxazole loss was highest in the microaerobic reactors; and phenytoin was recalcitrant in all reactors. Transformation products of sulfamethoxazole and desvenlafaxine resulted in the reformation...

Modelling the chemical behaviour of tellurium species in the reactor pressure vessel and the reactor cooling system under severe accident conditions

International Nuclear Information System (INIS)

Alonso, A.; Gonzalez, C.

1991-07-01

This state of the art report contains information on the behaviour of tellurium and its compounds in the reactor pressure vessel and the reactor coolant system under light water reactor severe accident conditions. To characterise tellurium behaviour, it is necessary the previous knowledge of the species of tellurium released from the core, and simultaneity of its release with that of other materials which can alter the transport, for instance, control rod and structural materials. Release and transport experiments have been reviewed along with the models implemented in the codes which are used in the international community: TRAPMELT, RAFT, VICTORIA and SOPHIE. From the experiments, it can be concluded that other species different to Te 2 , such as tin telluride and cesium telluride, may be released from the fuel. That is why they must be considered in the transport phenomena. There is also experimental evidence of the strong interaction of Te 2 with Inconel 600 and stainless steel of the pipe walls and structures, however this strong interaction is in competition with the interaction of tellurium with aerosols, which under severe accident conditions may represent an area greater than that of the primary system. It is for the absence of significant tellurium species in the transport models, and also for the interaction of tellurium with aerosols, for which some codes show the greatest deficiencies

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

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

1994-01-01

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

Generation of net electric power with a tokamak reactor under foreseeable physical and engineering conditions

International Nuclear Information System (INIS)

Hiwatari, R.; Asaoka, Y.; Okano, K.; Yoshida, T.; Tomabechi, K.

2004-01-01

This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with β N = 1.8, HH = 1.0, andf nGW 0.85 had sufficient potential to achieve the electric break-even condition (net electric power P e net = 0MW) under the following engineering conditions: machine major radius 6.5m ≤ R p ≤ 8.5m, the maximum magnetic field on TF coils B tmax = 16 T, thermal efficiency η e 30%, and NBI system efficiency η NBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are β N and fη GW . ≥ 3.0 is required for P e net ∼ 600MW with fusion power P f ∼ 3000MW. On the other hand, fη GW ≥ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of T ave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs P f (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius R p ∼ 7.5m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy. (author)

Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

Energy Technology Data Exchange (ETDEWEB)

Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

2014-01-15

Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

Reactor coolant pump shaft seal behavior during blackout conditions

International Nuclear Information System (INIS)

Mings, W.J.

1985-01-01

The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

Features, present condition of development and future scope on the high temperature gas reactor as an innovative one

International Nuclear Information System (INIS)

Shiozawa, Shusaku

2001-01-01

The high temperature gas reactor has some features without previous reactors such as high temperature capable of taking-out, high specific safety, feasibility adaptable to versatile fuel cycle, and so on. Then, it is expected to be an innovative reactor to contribute to diversification of energy supply and expansion of energy application field. In Japan, under the HTTR (high temperature engineering test reactor) plan, construction of HTTR, which is the first high temperature gas reactor in Japan, was finished and its output upgrading test has been promoted. And, on the HTTR plan, together with promotion of full power operation, reactor performance tests, safety proof test, and so on, it is planned to carry out study on application of the high temperature heat such as hydrogen production and so on to aim to practise establishment and upgrading of technologies on high temperature gas reactor in Japan. Here were introduced features and present condition of development of the high temperature gas reactor as an innovative type reactor and described role and future scope in Japan. (G.K.)

Characterization, treatment and conditioning of radioactive graphite from decommissioning of nuclear reactors

International Nuclear Information System (INIS)

2006-09-01

Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-production reactors. It is used primarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. Worldwide, there are more than 230 000 tonnes of radioactive graphite which will eventually need to be managed as radioactive waste. Proper management of radioactive graphite waste requires complex planning and the implementation of several interrelated operations. There are two basic options for graphite waste management: (1) packaging of non-conditioned graphite waste with subsequent direct disposal of the waste packages, and (2) conditioning of graphite waste (principally either by incineration or calcination) with separate disposal of any waste products produced, such as incinerator ash. In both cases, the specific properties of graphite - such as Wigner energy, graphite dust explosibility, and radioactive gases released from waste graphite - have a potential impact on the safety of radioactive graphite waste management and need to be carefully considered. Radioactive graphite waste management is not specifically addressed in IAEA publications. Only general and limited information is available in publications dealing with decommissioning of nuclear reactors. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle. The first draft report was prepared at a meeting on 23-27 February 1998. A technical meeting (TM) was held in October 1999 in coincidence with the Seminar on

Catalytic effect of different reactor materials under subcritical water conditions: decarboxylation of cysteic acid into taurine

Science.gov (United States)

Faisal, M.

2018-03-01

In order to understand the influence of reactor materials on the catalytic effect for a particular reaction, the decomposition of cysteic acid from Ni/Fe-based alloy reactors under subcritical water conditions was examined. Experiments were carried out in three batch reactors made of Inconel 625, Hastelloy C-22 and SUS 316 over temperatures of 200 to 300 °C. The highest amount of eluted metals was found for SUS 316. The results demonstrated that reactor materials contribute to the resulting product. Under the tested conditions, cysteic acid decomposes readily with SUS 316. However, the Ni-based materials (Inconel 625 and Hastelloy C-22) show better resistance to metal elution. It was found that among the materials used in this work, SUS 316 gave the highest reaction rate constant of 0.1934 s‑1. The same results were obtained at temperatures of 260 and 300 °C. Investigation of the Arrhenius activation energy revealed that the highest activation energy was for Hastelloy C-22 (109 kJ/mol), followed by Inconel 625 (90 kJ/mol) and SUS 316 (70 kJ/mol). The decomposition rate of cysteic acid was found to follow the results for the trend of the eluted metals. Therefore, it can be concluded that the decomposition of cysteic acid was catalyzed by the elution of heavy metals from the surface of the reactor. The highest amount of taurine from the decarboxylation of cysteic acid was obtained from SUS 316.

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

Seismic verification of the Italian PEC fast reactor and effects of seismic conditions on the design

International Nuclear Information System (INIS)

Martelli, A.; Cecchini, F.; Masoni, P.; Maresca, G.; Castoldi, A.

1988-01-01

This paper deals with the aseismic design features of the Italian PEC fast reactor and the effects of seismic conditions on the reactor design. More precisely, after some notes on the main plant features, the paper reports on the design earthquakes adopted, the seismic monitoring procedures and the related actions, the design requirements, criteria and methods, and also provides a brief summary of the main research and development studies performed in support of design analysis. For the above-mentioned items, comparisons with the other fast reactors of the European Community countries are presented. Furthermore, the paper stresses the design modifications adopted to guarantee PEC seismic safety

Conditioning CANDU reactor wastes for disposal

International Nuclear Information System (INIS)

Beamer, N.V.; Bourns, W.T.; Buckley, L.P.; Speranzini, R.A.

1981-12-01

A Waste Treatment Centre (WTC) is being constructed at the Chalk River Nuclear Laboratories to develop and demonstrate processes for converting reactor wastes to a form suitable for disposal. The WTC contains a starved air incinerator for reducing the volume of combustible solid wastes, a reverse osmosis section for reducing the volume of liquid wastes and an immobilization section for incorporating the conditioned wastes in bitumen. The incinerator is commissioned on inactive waste: approximately 16.5 Mg of waste packaged in polyethylene bags has been incinerated in 17 burns. Average weight and volume reductions of 8.4:1 and 32:1, respectively, have been achieved. Construction of the reverse osmosis section of WTC is complete and inactive commissioning will begin in 1982 January. The reverse osmosis section was designed to process 30,000 m 3 /a of dilute radioactive waste. The incinerator ash and concentrated aqueous waste will be immobiblized in bitumen using a horizontal mixer and wiped-film evaporator. Results obtained during inactive commissioning of the incinerator are described along with recent results of laboratory programs directed at demonstrating the reverse osmosis and bituminization processes

Advanced development and operating experience with a canned motor pump under pressurized water reactor conditions

International Nuclear Information System (INIS)

Dittmer, H.; Reymann, A.; Seibig, B.; Reinecker, H.

1988-01-01

At the research reactor FRG-2, Geesthacht, an irradiation device is in operation for testing defective light-water-reactor (LWR) test fuel rods under pressurized water reactor conditions (320 0 C, 160 bar). The requirements to the canned motor pump for cooling water circulation: medium: Demineralized water, operating temperature 320 0 C, operating pressure 155 bar, radiation field of the reactor, integration in the irradiation capsule, helium leak rate -6 mbar.dm 3 .s -1 , minimum working life 3000 hours, were high and caused difficulties in the acquisition of this component. First test runs with supplied pumps showed that the desired working life could not be achieved. The results of the development steps, the test runs, and the performance in service show that for our range of applications, the best combination of materials for the radial bearings is silicon-infiltrated SiC (8% free Si) against the same material. These bearings allowed a good working life for the pump to be achieved. (orig./GL) [de

Study on vortex cavitation in a compact fast reactor. Effects of system pressure on inception condition

International Nuclear Information System (INIS)

Hiroyuki Sato; Toshiki Ezure; Hideki Kamide

2005-01-01

A compact sodium reactor is designed as a commercialized fast reactor cycle system. A 1/10 scaled water experiment was performed to optimize flow in an upper plenum of the reactor vessel, because of high flow velocity resulted from the compacted vessel. In the experiment, vortex cavitation was found at the hot leg inlet because of high velocity in the hot leg pipe (9.4m/s in the design). To evaluate cavitation inception condition of the commercialized reactor, we use the cavitation number k in order to consider the difference of system pressures (0.1MPa in the experiment and 0.3MPa in the design). The minimum pressure at the vortex center will depend on vortex core radius (size of forced vortex region). It is related to axial velocity gradient and fluid viscosity in theory of the Burger's stretched vortex model. We carried out a basic water experiment to investigate the influence of system pressure and fluid viscosity on the vortex cavitation. The cavitation number at the inception of vortex cavitation slightly increased according to the increase of the system pressure. It means that the vortex cavitation occurs easily under higher pressure condition as compared with the similar condition of cavitation number with lower pressure. However the increase was less than 30% when the system pressure was varied from 0.1 to 0.3MPa. The influence of fluid viscosity was examined by change of fluid temperature. Velocity distribution around the vortex was also measured to see the structure of vortex. (authors)

IAEA Launches Expert Advisory Service for Research Reactor Infrastructure, First Mission to Nigeria

International Nuclear Information System (INIS)

2018-01-01

The International Atomic Energy Agency (IAEA) has launched a new peer review service to assist Member States in the development of infrastructure for nuclear research reactors, expanding the range of its expert advisory missions. The first Integrated Nuclear Infrastructure Review for Research Reactors (INIR-RR) was conducted this week in Nigeria at the invitation of the Government, which is planning to build the country’s second research reactor. Research reactors are used for research, development, education and training. They play a vital role across several fields, producing radioisotopes used in research, medicine, industry and agriculture. Operation of a research reactor requires a national infrastructure — including a legal and regulatory framework — to ensure that national and international obligations are met during planning, design, construction, operation and decommissioning.

Study of startup conditions of a pulsed annular reactor; Estudo das reacoes de partida de um reator anelar pulsado

Energy Technology Data Exchange (ETDEWEB)

Silva, Mario Augusto Bezerra da

2003-10-15

A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

Design report on the guide box-reactivity and safety control plates for MPR reactor under normal operation conditions

International Nuclear Information System (INIS)

Markiewicz, M.

1999-01-01

The reactivity control system for the MPR reactor (Multi Purpose Reactor) is a critical component regarding safety, it must ensure a fast shut down, maintaining the reactor in subcritical condition under normal or accidental operation condition. For this purpose, this core component must be designed to maintain its operating capacity during all the residence time and under any foreseen operation condition. The mechanical design of control plates and guide boxes must comply with structural integrity, maintaining its geometric and dimensional stability within the pre-established limits to prevent interferences with other core components. For this, the heat generation effect, mechanical loads and environment and irradiation effects were evaluated during the mechanical design. The reactivity control system is composed of guide boxes, manufactured from Aluminium alloy, located between the fuel elements, and control absorber plates of Ag-In-Cd alloy hermetically enclosed by a cladding of stainless steel sliding inside de guide boxes. The upward-downward movement is transmitted by a rod from the motion device located at the reactor lower part. The design requirements, criteria and limits were established to fulfill with the normal and abnormal operation conditions. The design verifications were performed by analytical method, estimating the guide box and control plates residence time. The result of the analysis performed, shows that the design of the reactivity control system and the material selected, are appropriate to fulfill the functional requirements, with no failures attributed to the mechanical design. (author)

Decay heat removal and transient analysis in accidental conditions in the EFIT reactor

International Nuclear Information System (INIS)

Bandini, G.; Meloni, P.; Polidori, M.; Casamirra, M.; Castiglia, F.; Giardina, M.

2007-01-01

The development of a conceptual design of an industrial scale transmutation facility (EFIT) of several 100 MW thermal power based on Accelerator Driven System (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which lead to the Loss of Heat Sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1-D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios. (author)

Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

Directory of Open Access Journals (Sweden)

Giacomino Bandini

2008-01-01

Full Text Available The development of a conceptual design of an industrial-scale transmutation facility (EFIT of several 100 MW thermal power based on accelerator-driven system (ADS is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS. In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

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.)

Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

Energy Technology Data Exchange (ETDEWEB)

Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain); Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland)

2004-07-01

One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

International Nuclear Information System (INIS)

Ernestova, M.; Zamboch, M.; Devrient, B.; Roth, A.; Ehrnsten, U.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ritter, S.; Seifert, H.P.

2004-01-01

One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

2016-12-15

Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

International Nuclear Information System (INIS)

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

2016-01-01

Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

Test of safety injection supply by diesel generator under reactor vessel closed condition

International Nuclear Information System (INIS)

Zhang Hao; Bi Fengchuan; Che Junxia; Zhang Jianwen; Yang Bo

2014-01-01

The paper studied that the test of diesel generator full load take-up under the condition of actual safety injection and reactor vessel closed in Ningde nuclear project unit l. It is proved that test result accorded with design criteria, meanwhile, the test was removed from the key path of project schedule, which cut a huge cost. (authors)

Requirements for light water reactors

International Nuclear Information System (INIS)

Hedin, F.

2009-01-01

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

Propagation of cracks by stress corrosion in conditions of BWR type reactor

International Nuclear Information System (INIS)

Merino C, F.J.; Fuentes C, P.

2004-01-01

In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

Biogas Upgrading via Hydrogenotrophic Methanogenesis in Two-Stage Continuous Stirred Tank Reactors at Mesophilic and Thermophilic Conditions

DEFF Research Database (Denmark)

Bassani, Ilaria; Kougias, Panagiotis; Treu, Laura

2015-01-01

This study proposes an innovative setup composed by two stage reactors to achieve biogas upgrading coupling the CO2 in the biogas with external H2 and subsequent conversion into CH4 by hydrogenotrophic methanogenesis. In this configuration, the biogas produced in the first reactor was transferred...... production and CO2 conversion was recorded. The consequent increase of pH did not inhibit the process indicating adaptation of microorganisms to higher pH levels. The effects of H2 on the microbial community were studied using high-throughput Illumina random sequences and full-length 16S rRNA genes extracted...... to the second one, where H2 was injected. This configuration was tested at both mesophilic and thermophilic conditions. After H2 addition, the produced biogas was upgraded to average CH4 content of 89% in the mesophilic reactor and 85% in the thermophilic. At thermophilic conditions, a higher efficiency of CH4...

Formation of metabolites during biodegradation of linear alkylbenzene sulfonate in an upflow anaerobic sludge bed reactor under thermophilic conditions

DEFF Research Database (Denmark)

Mogensen, Anders Skibsted; Ahring, Birgitte Kiær

2002-01-01

Biodegradation of linear alkylbenzene sulfonate (LAS) was shown in an upflow anaerobic sludge blanket reactor under thermophilic conditions. The reactor was inoculated with granular biomass and fed with a synthetic medium and 3 mumol/L of a mixture of LAS with alkylchain length of 10 to 13 carbon...

Stabilization of burn conditions in a thermonuclear reactor using artificial neural networks

Science.gov (United States)

Vitela, Javier E.; Martinell, Julio J.

1998-02-01

In this work we develop an artificial neural network (ANN) for the feedback stabilization of a thermonuclear reactor at nearly ignited burn conditions. A volume-averaged zero-dimensional nonlinear model is used to represent the time evolution of the electron density, the relative density of alpha particles and the temperature of the plasma, where a particular scaling law for the energy confinement time previously used by other authors, was adopted. The control actions include the concurrent modulation of the D-T refuelling rate, the injection of a neutral He-4 beam and an auxiliary heating power modulation, which are constrained to take values within a maximum and minimum levels. For this purpose a feedforward multilayer artificial neural network with sigmoidal activation function is trained using a back-propagation through-time technique. Numerical examples are used to illustrate the behaviour of the resulting ANN-dynamical system configuration. It is concluded that the resulting ANN can successfully stabilize the nonlinear model of the thermonuclear reactor at nearly ignited conditions for temperature and density departures significantly far from their nominal operating values. The NN-dynamical system configuration is shown to be robust with respect to the thermalization time of the alpha particles for perturbations within the region used to train the NN.

Stabilization of burn conditions in a thermonuclear reactor using artificial neural networks

International Nuclear Information System (INIS)

Vitela, J.E.; Martinell, J.J.

1998-01-01

In this work we develop an artificial neural network (ANN) for the feedback stabilization of a thermonuclear reactor at nearly ignited burn conditions. A volume-averaged zero-dimensional nonlinear model is used to represent the time evolution of the electron density, the relative density of alpha particles and the temperature of the plasma, where a particular scaling law for the energy confinement time previously used by other authors, was adopted. The control actions include the concurrent modulation of the D-T refuelling rate, the injection of a neutral He-4 beam and an auxiliary heating power modulation, which are constrained to take values within a maximum and minimum levels. For this purpose a feedforward multilayer artificial neural network with sigmoidal activation function is trained using a back-propagation through-time technique. Numerical examples are used to illustrate the behaviour of the resulting ANN-dynamical system configuration. It is concluded that the resulting ANN can successfully stabilize the nonlinear model of the thermonuclear reactor at nearly ignited conditions for temperature and density departures significantly far from their nominal operating values. The NN-dynamical system configuration is shown to be robust with respect to the thermalization time of the alpha particles for perturbations within the region used to train the NN. (author)

Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

1986-10-01

We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

1995-09-01

Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

International Nuclear Information System (INIS)

Kulikov, I.S.; Tverkovkin, B.E.

1984-01-01

The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

International Nuclear Information System (INIS)

Harris, D.R.

1995-01-01

Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

Energy Technology Data Exchange (ETDEWEB)

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

On the influence of fusion reactor conditions on optical properties of metallic plasma-viewing mirrors

International Nuclear Information System (INIS)

Voitsenya, V.S.; Gritsyna, V.I.; Konovalov, V.G.; Ruzhitskij, V.V.; Shapoval, A.N.; Orlinskij, D.V.

1997-01-01

This paper presents the results of imitation experiments concerning the effects of fusion reactor conditions on the properties of mirrors made of stainless steel, copper and beryllium. The neutron irradiation was imitated using MeV energy range ions. To imitate the effects of charge exchange atoms (CXA) bombardment, keV energy range D + and He + ions were used. From the data obtained it was concluded that not only the reflectivity but also the resistance to CXA sputtering have to be taken into account when choosing the materials for the first mirrors of a fusion reactor. (orig.)

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Adaptation of continuous biogas reactors operating under wet fermentation conditions to dry conditions with corn stover as substrate.

Science.gov (United States)

Kakuk, Balázs; Kovács, Kornél L; Szuhaj, Márk; Rákhely, Gábor; Bagi, Zoltán

2017-08-01

Corn stover (CS) is the agricultural by-product of maize cultivation. Due to its high abundance and high energy content it is a promising substrate for the bioenergy sector. However, it is currently neglected in industrial scale biogas plants, because of its slow decomposition and hydrophobic character. To assess the maximum biomethane potential of CS, long-term batch fermentations were carried out with various substrate concentrations and particle sizes for 72 days. In separate experiments we adapted the biogas producing microbial community in wet fermentation arrangement first to the lignocellulosic substrate, in Continuous Stirred Tank Reactor (CSTR), then subsequently, by continuously elevating the feed-in concentration, to dry conditions in solid state fermenters (SS-AD). In the batch tests, the CSTR experiment, the daily substrate loading was gradually increased from 1 to 2 g vs /L/day until the system produced signs of overloading. Then the biomass was transferred to SS-AD reactors and the adaptation process was studied. Although the specific methane yields were lower in the SS-AD arrangement (177 mL CH 4 /g vs in CSTR vs. 105 mL in SS-AD), the benefits of process operational parameters, i.e. lower energy consumption, smaller reactor volume, digestate amount generated and simpler configuration, may compensate the somewhat lower yield. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

Proceedings [of the] symposium on zirconium alloys for reactor components

International Nuclear Information System (INIS)

1992-01-01

A two day symposium on zirconium alloys for reactor components (ZARC-91) was organised during 12-13, 1991. There were 6 invited talks and 43 contributed papers in 10 technical sessions. This symposium, took stock of the progress achieved in the development, design, fabrication and quality assurance of zirconium alloy components and emphasized the R and D efforts required for meeting the challenges posed by the rapid growth of nuclear power in our country. Topics like physical metallurgy, corrosion and irradiation behaviour, and in-service inspection were also covered. The proceedings/papers are arranged under the headings: (1)invited talks, (2)fabrication, (3)design requirement, (4)quality assurance, (5)irradiation damage and PIE, (6)corrosion and hydriding, and (7)in-service inspection. (N.B.). refs., figs., tabs

Indian Chemical Engineering Congress 1995: 48th annual session of Indian Institute of Chemical Engineers: abstracts and invited lectures

International Nuclear Information System (INIS)

1995-01-01

The 48th Annual Session of Indian Institute of Chemical Engineers was held in Kalpakkam during December 27-30, 1995. The book contains the proceeding of the conference, both abstracts and invited lectures. The topics covered included various aspects pertaining to chemical engineering and technology along with the chemical and engineering processes relevant to nuclear fuel cycle like uranium ore processing, fuel fabrication, reactor operation, fuel reprocessing and radioactive waste management. Papers relevant to INIS are indexed separately

Fluidized-bed reactors processes and operating conditions

CERN Document Server

Yates, John G

2016-01-01

The fluidized-bed reactor is the centerpiece of industrial fluidization processes. This book focuses on the design and operation of fluidized beds in many different industrial processes, emphasizing the rationale for choosing fluidized beds for each particular process. The book starts with a brief history of fluidization from its inception in the 1940’s. The authors present both the fluid dynamics of gas-solid fluidized beds and the extensive experimental studies of operating systems and they set them in the context of operating processes that use fluid-bed reactors. Chemical engineering students and postdocs as well as practicing engineers will find great interest in this book.

Effect of decontamination on oxidation of austenitic stainless steel in reactor conditions

International Nuclear Information System (INIS)

Starkman, T.

1984-07-01

Austenitic stainless steels were oxidized in static autoclaves in light water reactor conditions. After the autoclave treatments the specimens were decontaminated with the aid of alkaline potassium permanganate (AP) and oxalic and citric acid (CITROX) as well as electrochemically in H 3 PO 4 . Alternating oxidation and decontamination tests were performed. An elemental analysis of the surfaces of the specimens was carried out by electron spectroscopy. Changes in structures and thicknesses of the oxide layers were observed. (author)

A structural evaluation of the Shippingport reactor pressure vessel for transport impact conditions

International Nuclear Information System (INIS)

Witte, M.C.; Chou, C.K.

1989-01-01

The Shippingport Atomic Power Station in Shippingport, Pennsylvania, is being decommissioned and dismantled. This government-leased property will be returned, in a radiologically safe condition, to its owner. All radioactive material is being removed from the Shippingport Station and transported for burial to the DOE Hanford Reservation in Richland, Washington. The reactor pressure vessel (RPV) will be transported by barge to Hanford. This paper describes an evaluation of the structural response of the RPV to the normal and accident impact test conditions as required by the Code of Federal Regulations. 3 refs., 5 figs., 3 tabs

Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

International Nuclear Information System (INIS)

Fuentes C, P.

2003-01-01

This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

Effect of inlet conditions on the performance of a palladium membrane reactor

International Nuclear Information System (INIS)

Birdsell, S.A.; Willms, R.S.; Arzu, P.; Costello, A.

1997-10-01

Palladium membrane reactors (PMR) will be used to remove tritium and other hydrogen isotopes from impurities, such as tritiated methane and tritiated water, in the exhaust of the International Thermonuclear Experimental Reactor. In addition to fusion-fuel processing, the PMR system can be used to recover tritium from tritiated waste water. This paper investigates the effect of inlet conditions on the performance of a PMR. A set of experiments were run to determine, independently, the effect of inlet compositions and residence time on performance. Also, the experiments were designed to determine if the injected form of hydrogen (CH 4 or H 2 O) effects performance. Results show that the PMR operates at optimal hydrogen recovery with a broad range of inlet compositions and performance is shown to increase with increased residence time. PMR performance is shown to be independent of whether hydrogen is injected in the form of CH 4 or H 2 O

Bioremediation of polychlorinated-p-dioxins/dibenzofurans contaminated soil using simulated compost-amended landfill reactors under hypoxic conditions

Energy Technology Data Exchange (ETDEWEB)

Chen, Wei-Yu; Wu, Jer-Horng, E-mail: enewujh@mail.ncku.edu.tw; Lin, Shih-Chiang; Chang, Juu-En

2016-07-15

Highlights: • We developed a new hypoxic reactor system for remediating PCDD/Fs. • We demonstrated effects of compost on the degradation of PCDD/Fs. • We uncovered microbial compositions and dynamics during the degradation of PCDD/Fs. - Abstract: Compost-amended landfill reactors were developed to reduce polychlorinated-p-dioxins and dibenzofurans (PCDD/Fs) in contaminated soils. By periodically recirculating leachate and suppling oxygen, the online monitoring of the oxidation reduction potential confirmed that the reactors were maintained under hypoxic conditions, with redox levels constantly fluctuating between −400 and +80 mV. The subsequent reactor operation demonstrated that PCDD/F degradation in soil could be facilitated by amending compost originating from the cow manure and waste sludge and that the degradation might be affected by the availability of easily degradable substrates in the soil and compost. The pyrosequencing analysis of V4/V5 regions of bacterial 16S rRNA genes suggested that species richness of the soil microbial community was increased by a factor of 1.37–1.61. Although the bacterial community varied with the compost origin and changed markedly during reactor operation, it was dominated by Alphaproteobacteria, Gammaproteobacteria, Actinobacteria, and Firmicutes. The aerotolerant anaerobic Sedimentibacter and Propionibacterium spp., and the uncultured Chloroflexi group could be temporarily induced to a high abundance by amending the cow manure compost; the bacterial growths were associated with the rapid degradation of PCDD/Fs. Overall, the novel bioremediation method for PCDD/F-contaminated soils using hypoxic conditions was effective, simple, energy saving, and thus easily practicable.

Bioremediation of polychlorinated-p-dioxins/dibenzofurans contaminated soil using simulated compost-amended landfill reactors under hypoxic conditions

International Nuclear Information System (INIS)

Chen, Wei-Yu; Wu, Jer-Horng; Lin, Shih-Chiang; Chang, Juu-En

2016-01-01

Highlights: • We developed a new hypoxic reactor system for remediating PCDD/Fs. • We demonstrated effects of compost on the degradation of PCDD/Fs. • We uncovered microbial compositions and dynamics during the degradation of PCDD/Fs. - Abstract: Compost-amended landfill reactors were developed to reduce polychlorinated-p-dioxins and dibenzofurans (PCDD/Fs) in contaminated soils. By periodically recirculating leachate and suppling oxygen, the online monitoring of the oxidation reduction potential confirmed that the reactors were maintained under hypoxic conditions, with redox levels constantly fluctuating between −400 and +80 mV. The subsequent reactor operation demonstrated that PCDD/F degradation in soil could be facilitated by amending compost originating from the cow manure and waste sludge and that the degradation might be affected by the availability of easily degradable substrates in the soil and compost. The pyrosequencing analysis of V4/V5 regions of bacterial 16S rRNA genes suggested that species richness of the soil microbial community was increased by a factor of 1.37–1.61. Although the bacterial community varied with the compost origin and changed markedly during reactor operation, it was dominated by Alphaproteobacteria, Gammaproteobacteria, Actinobacteria, and Firmicutes. The aerotolerant anaerobic Sedimentibacter and Propionibacterium spp., and the uncultured Chloroflexi group could be temporarily induced to a high abundance by amending the cow manure compost; the bacterial growths were associated with the rapid degradation of PCDD/Fs. Overall, the novel bioremediation method for PCDD/F-contaminated soils using hypoxic conditions was effective, simple, energy saving, and thus easily practicable.

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

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

Experimental and Computational Study of the Hydrodynamics of Trickle Bed Flow Reactor Operating Under Different Pressure Conditions

Science.gov (United States)

Rabbani, S.; Ben Salem, I.; Nadeem, H.; Kurnia, J. C.; Shamim, T.; Sassi, M.

2014-12-01

Pressure drop estimation and prediction of liquid holdup play a crucial role in design and operation of trickle bed reactors. Experiments are performed for Light Gas Oil (LGO)-nitrogen system in ambient temperature conditions in an industrial pilot plant with reactor height 0.79 m and diameter of 0.0183 m and pressure ranging from atmospheric to 10 bars. It was found that pressure drop increased with increase in system pressure, superficial gas velocity and superficial liquid velocity. It was demonstrated in the experiments that liquid holdup of the system increases with the increase in superficial liquid velocity and tends to decrease with increase in superficial gas velocity which is in good agreement with existing literature. Similar conditions were also simulated using CFD-software FLUENT. The Volume of Fluid (VoF) technique was employed in combination with "discrete particle approach" and results were compared with that of experiments. The overall pressure drop results were compared with the different available models and a new comprehensive model was proposed to predict the pressure drop in Trickle Bed Flow Reactor.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. 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 Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. 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 Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Limits and conditions for continuous operation of WWR-S reactor

International Nuclear Information System (INIS)

Pittermann, P.; Listik, E.

1979-02-01

The fundamental technological and nuclear characteristics of the WWR-S reactor, safety limits and concepts of technical surveillance with particular attention to radiation safety of staff and of neighbouring population are outlined. The rules are mandatory for the reactor staff and for the users. The material is part of safety documentation for the WWR-S reactor. (author)

The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

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.)

Development of a diagnostic system for identifying accident conditions in a reactor

International Nuclear Information System (INIS)

Santhosh; Gera, B.; Kumar, Mithilesh; Thangamani, I.; Prasad, Hari; Srivastava, A.; Dutta, Anu; Sharma, Pavan K.; Majumdar, P.; Verma, V.; Mukhopadhyay, D.; Ganju, Sunil; Chatterjee, B.; Sanyasi Rao, V.V.S.; Lele, H.G.; Ghosh, A.K.

2009-07-01

This report describes a methodology for identification of accident conditions in a nuclear reactor from the signals available to the operator. A large database of such signals is generated through analyses - for core, containment, environmental dispersion and radiological dose to train a computer code based on an Artificial Neural Networks (ANNs). At present, in the prediction mode, information on LOCA (location and size of break), status of availability of ECCS, and expected doses can be predicted well for a 220 MWe PHWR. (author)

Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

International Nuclear Information System (INIS)

Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

1996-12-01

Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

Bearings with a key role in nuclear reactor safety

International Nuclear Information System (INIS)

Jackson, J.

1981-01-01

The most critical period during the operation of heavy-duty pumps occurs at run-up, when sudden high loads coincide with below normal oil pressure. When those pumps are a key component of a nuclear reactor safety-system, reliability takes on a special meaning. Electricite de France recently invited a British company, Michell Bearings, to find a solution to these 'black-start' problems, the details of which are described in the following article

Safety analysis of switching between reductive and oxidative conditions in a reaction coupling reverse flow reactor.

NARCIS (Netherlands)

van Sint Annaland, M.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

2001-01-01

A new reverse flow reactor is developed where endothermic reactants (propane dehydrogenation) and exothermic reactants (fuel combustion) are fed sequentially to a monolithic catalyst, while periodically alternating the inlet and outlet positions. Upon switching from reductive to oxidative conditions

Effects of ocean conditions upon the passive residual heat removal system (PRHRS) of ship reactor

International Nuclear Information System (INIS)

Su Guanghui; Zhang Jinling; Guo Yujun; Qiu Suizheng; Yu Zhenwan; Jia Dounan

1996-01-01

The authors investigate the influence of ocean conditions (heaving, listing, rolling) on the natural circulation flow and the ability of heat transfer of the ship reactor's PRHRS, and develops a mathematical model. A program, MISAP 02, is compiled with the structured FORTRAN 77 using the advanced Gear method. the program is used to calculate the above influence. The results show that the ocean conditions have some effects on the natural circulation flow and the ability of heat transfer

STUDY OF MERCURY OXIDATION BY SCR CATALYST IN AN ENTRAINED-FLOW REACTOR UNDER SIMULATED PRB CONDITIONS

Science.gov (United States)

A bench-scale entrained-flow reactor system was constructed for studying elemental mercury oxidation under selective catalytic reduction (SCR) reaction conditions. Simulated flue gas was doped with fly ash collected from a subbituminous Powder River Basin (PRB) coal-fired boiler ...

An invitation to critical mathematics education

DEFF Research Database (Denmark)

Skovsmose, Ole

An Invitation to Critical Mathematics Education deals with a range of crucial topics. Among these are students’ foreground, landscapes of investigation, and mathematics in action. The book is intended for a broad audience: educators, students, teachers, policy makers, anybody interested...... in the further development of mathematics education. The book discusses concerns and preoccupation. This way it provides an invitation into critical mathematics education....

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

International Nuclear Information System (INIS)

Kobayashi, Michiyuki; Murata, Hiroyuki; Sawada, Kenichi; Inasaka, Fujio; Aya, Izuo; Shiozaki, Koki

1999-01-01

By inputting the experimental data, information and others on thermo-hydraulic characteristics of integrated ship propulsion reactor accumulated hitherto by the Ship Research Institute and some recent cooperation results into the nuclear ship engineering simulation system, it was conducted not only to contribute an improvement study on next ship reactor by executing general analysis and evaluation on motion characteristics under ship body motion conditions, safety at accidents, and others of the integrated ship reactor but also to investigate and prepare some measures to apply fundamental experiment results based on obtained here information to safety countermeasure of the nuclear ships. In 1997 fiscal year, on safety of the integrated ship propulsion reactor loading nuclear ship, by adding experimental data on unstable flow analysis and information on all around of the analysis to general data base fundamental program, development to intellectual data base program was intended; on effect of pulsation flow on thermo-hydraulic characteristics of ship propulsion reactor; after pulsation flow visualization experiment, experimental equipment was reconstructed into heat transfer type to conduct numerical analysis of pulsation flow by confirming validity of numerical analysis code under comparison with the visualization experiment results; and on thermo-hydraulic behavior in storage container at accident of active safety type ship propulsion reactor; a flashing vibration test using new apparatus finished on its higher pressurization at last fiscal year to examine effects of each parameter such as radius and length of exhausting nozzle and pool water temperature. (G.K.)

WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS; A

International Nuclear Information System (INIS)

Carl R.F. Lund

2001-01-01

This report covers the second year of a project investigating water-gas shift catalysts for use in membrane reactors. It has been established that a simple iron high temperature shift catalyst becomes ineffective in a membrane reactor because the reaction rate is severely inhibited by the build-up of the product CO(sub 2). During the past year, an improved microkinetic model for water-gas shift over iron oxide was developed. Its principal advantage over prior models is that it displays the correct asymptotic behavior at all temperatures and pressures as the composition approaches equilibrium. This model has been used to explore whether it might be possible to improve the performance of iron high temperature shift catalysts under conditions of high CO(sub 2) partial pressure. The model predicts that weakening the surface oxygen bond strength by less than 5% should lead to higher catalytic activity as well as resistance to rate inhibition at higher CO(sub 2) partial pressures. Two promoted iron high temperature shift catalysts were studied. Ceria and copper were each studied as promoters since there were indications in the literature that they might weaken the surface oxygen bond strength. Ceria was found to be ineffective as a promoter, but preliminary results with copper promoted FeCr high temperature shift catalyst show it to be much more resistant to rate inhibition by high levels of CO(sub 2). Finally, the performance of sulfided CoMo/Al(sub 2)O(sub 3) catalysts under conditions of high CO(sub 2) partial pressure was simulated using an available microkinetic model for water-gas shift over this catalyst. The model suggests that this catalyst might be quite effective in a medium temperature water-gas shift membrane reactor, provided that the membrane was resistant to the H(sub 2)S that is required in the feed

Transactions of the twenty-fifth water reactor safety information meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

Transactions of the twenty-fifth water reactor safety information meeting

International Nuclear Information System (INIS)

Monteleone, S.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session

Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

International Nuclear Information System (INIS)

Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

2015-01-01

During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

Energy Technology Data Exchange (ETDEWEB)

Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Van Tichelen, Katrien [SCK- CEN, Boeretang 200, 2400 Mol (Belgium)

2017-02-15

Highlights: • Quantification of the design and safety of the MYRRHA reactor in the event of a pin failure. • Simulation of different accident scenarios in both forced and natural convection regime. • The accumulation areas at the free-surface in case of the least dense particles depend on the flow regime. • The densest particles form an important deposit at the bottom of the vessel. • Further study of the risk of core blockage requires a detailed model of the core. - Abstract: This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for

Nuclear reactor safety systems

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1980-01-01

A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

International Nuclear Information System (INIS)

Cruz, C.B.L.

1990-01-01

The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  Personal invitations for population-based breast cancer screening

DEFF Research Database (Denmark)

Saalasti-Koskinen, Ulla; Mäkelä, Marjukka; Saarenmaa, Irma

2010-01-01

participation free of charge and the benefits of detecting breast cancer early. Harm associated with screening was seldom mentioned; no unit mentioned the possibility of false-negative results or overtreatment. CONCLUSION: The screening units provided very variable information, which often was biased toward......RATIONALE AND OBJECTIVES: Women who are invited for breast cancer screening should get enough information about the benefits and harms of screening to make an informed decision on participation. Personal invitations are an important source of information, because all invited women receive them....... The objective of this study was to evaluate the information breast cancer screening units send to women invited for screening in Finland. MATERIALS AND METHODS: A questionnaire was sent to all breast cancer screening units in Finland in 2005 and 2008, and the information (eg, invitations, results letters...

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

International Nuclear Information System (INIS)

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

1990-05-01

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

System for stress corrosion conditions tests on PWR reactors

International Nuclear Information System (INIS)

Castro, Andre Cesar de Jesus

2007-01-01

The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

Analysis of radionuclide dispersion at normal condition for AEC 1000 MW reactor power

International Nuclear Information System (INIS)

Sri Kuntjoro

2010-01-01

Analysis for radionuclide dispersion for the Atomic Energy Agency (AEC) 3,568 MWth Power Reactor, equal to the 1,000 MWe at normal condition has been done. Analysis was done for two piles that is separated by 500 m distance and angle of 90° one to other. Initial pace in doing the analysis is to determine reactors source term using ORIGEN2 and EMERALD NORMAL. computer code program. ORIGEN2 applied to determine radionuclide inventory emerged in the reactor. Hereinafter, by using Emerald Normal Computer code is calculated source term reaching the reactor stack. To analyze dose received by population is done by using PC-CREAM computer code. Calculation done for one and two PLTN attached in site candidate of plants. The result showed is that the highest radionuclide release for one PLTN is at 1 km distance and to 9 th zone toward ( 19.25° ) and for two PLTN is at 1 km distance and to 10 th zone toward (21.75° ). Radionuclide which up to population through two pathways that are foodstuff and inhalation. To foodstuff comes from radionuclide I 131 , and the biggest passed from milk product with 53.40 % for one and also two PLTN For inhalation pathway the highest radionuclide contribution come from Kr 85m is about 53.80 %. The highest total dose received by population is at 1 Km distance received by baby that is 4.10 µSi and 11.26 µSi for one and two PLTN respectively. Those result are very small compared to the maximum permission dose to population issued by regulatory body that is equal to 1 mSi. (author)

Biological Phosphorus Release and Uptake Under Alternating Anaerobic and Anoxic Conditions In a Fixed-Film Reactor

DEFF Research Database (Denmark)

Kerrn-Jespersen, Jens Peter; Henze, Mogens; Strube, Rune

1994-01-01

Biological phosphorus removal was investigated in a fixed-film reactor with alternating anaerobic and anoxic conditions. The tests showed that biological phosphorus removal can be obtained in a fixed-film reactor with nitrate as oxidising agent. In the anaerobic period, 0.52 mg of PO4-P...... was released per mg of acetate taken up on an average. In the anoxic period, 2.0 mg of PO4-P was taken up per mg of NO3-N reduced on an average. The relationship between potassium released and phosphate released in the anaerobic phase was determined to be 0.37 mg K/mg P, while the relationship between...

Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

International Nuclear Information System (INIS)

Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

1983-01-01

A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

Trench reactor: an overview

International Nuclear Information System (INIS)

Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

1988-01-01

Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

Multiregion, multigroup collision probability method with white boundary condition for light water reactor thermalization calculations

International Nuclear Information System (INIS)

Ozgener, B.; Ozgener, H.A.

2005-01-01

A multiregion, multigroup collision probability method with white boundary condition is developed for thermalization calculations of light water moderated reactors. Hydrogen scatterings are treated by Nelkin's kernel while scatterings from other nuclei are assumed to obey the free-gas scattering kernel. The isotropic return (white) boundary condition is applied directly by using the appropriate collision probabilities. Comparisons with alternate numerical methods show the validity of the present formulation. Comparisons with some experimental results indicate that the present formulation is capable of calculating disadvantage factors which are closer to the experimental results than alternative methods

Cleaning and air conditioning device for atmosphere in thermonuclear reactor chamber

International Nuclear Information System (INIS)

Ishida, Seiji.

1993-01-01

The device of the present invention removes tritium efficiently and attains ventilation and conditioning of a great amount of air flow. That is, there are disposed a humidity separator, a filter, a heater, a catalyst filled layer, a water jetting type humidifying heat insulation cooler and a cooler in this order from an inlet side (upstream) of contaminated room atmospheric gases. The catalyst filled layer, etc. are incorporated integrally into the ventilation air conditioning facility for ventilating air in the chamber of the thermonuclear reactor, to clean a tritium atmosphere at the same time. Accordingly, the device is made compact as a whole. A limit for the air flow rate owing to the use of the conventional catalyst tower and adsorbing tower is eliminated. Then a ventilating air conditioning for a great flow rate can be attained. Tritium is removed by cooling and dehumidification without using any adsorbent. Accordingly, an adsorbing tower is no more necessary and conventional regeneration operation is not required. As a result, space for installation is reduced, the system is simplified and the cost for construction and facility can be reduced. (I.S.)

Thermal stress analysis of reactor containment building considering severe weather condition

International Nuclear Information System (INIS)

Lee, Yun; Kim, Yun-Yong; Hyun, Jung-Hwan; Kim, Do-Gyeum

2014-01-01

Highlights: • We examine that through-wall crack risk in cold weather is high. • It is predicted that cracking in concrete wall will not happen in hot region. • Cracking due to hydration heat can be controlled by appropriate curing condition. • Temperature differences between inner and outer face is relatively small in hot weather. - Abstract: Prediction of concrete cracking due to hydration heat in mass concrete such as reactor containment building (RCB) in nuclear power plant is a crucial issue in construction site. In this study, the numerical analysis for heat transfer and stress development is performed for the containment wall in RCB by considering the severe weather conditions. Finally, concrete cracking risk in hot and cold weather is discussed based on analysis results. In analyses considering severe weather conditions, it is found that the through-wall cracking risk in cold weather is high due to the abrupt temperature difference between inside concrete and the ambient air in cold region. In hot weather, temperature differences between inner and outer face is relatively small, and accordingly the relevant cracking risk is relatively low in contrast with cold weather

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Challenge, Confrontation, and Exhortation as Intentional Invitations by Professional Helpers.

Science.gov (United States)

Schmidt, John J.

1996-01-01

Examines intentional invitations that challenge, confront, exhort, and persuade people to change their behaviors. Assumes that the sender controls the "intention" and that the receiver determines the degree of "inviting." Suggests that elements of the invitational model serve as a framework to create acceptable inducements in…

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

Transactions of the twenty-second water reactor safety information meeting

International Nuclear Information System (INIS)

1994-10-01

This report contains summaries of papers on reactor safety research to be presented at the 22nd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 24--26, 1994. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session. Individual papers have been cataloged separately

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

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

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

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

Effects of Personalization and Invitation Email Length on Web-Based Survey Response Rates

Science.gov (United States)

Trespalacios, Jesús H.; Perkins, Ross A.

2016-01-01

Individual strategies to increase response rate and survey completion have been extensively researched. Recently, efforts have been made to investigate a combination of interventions to yield better response rates for web-based surveys. This study examined the effects of four different survey invitation conditions on response rate. From a large…

A Suggestion for Restoring Public Support for Public Schools: Learn to Invite Patrons.

Science.gov (United States)

Roebuck, Flora N.; Russell, Gene; Wedel, Mack

1998-01-01

Proposes that the increasing disavowal of public schools results from experience with day-by-day disinviting behaviors. The suggested remedy is for educators to evaluate their on-the-job responses to their patrons. Wherever subminimal interpersonal conditions are identified, participants should learn to invite people to education. Presents a case…

Fast breeder reactors: Experience and trends. V. 2

International Nuclear Information System (INIS)

1986-01-01

The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

Evaluation of the Community's nuclear reactor safety research programme

International Nuclear Information System (INIS)

Brandstetter, A.; Goedkoop, J.A.; Jaumotte, A.; Malhouitre, G.; Tomkins, B.; Zorzoli, G.B.

1986-01-01

This report describes an evaluation of the 1980-85 CEC reactor safety programme prepared, at the invitation of the Commission, by a panel of six independent experts by means of examining the relevant document and by holding hearings with the responsible CEC staff. It contains the recommendations made by the panel on the following topics: the need for the JRC to continue to make its competence in the reactor safety field available to the Community; the importance of continuity in the JRC and shared-cost action programmes; the difficulty of developing reactor safety research programmes which satisfy the needs of users with diverse needs; the monitoring of the utilization of the research results; the maintenance of the JRC computer codes used by the Member States; the spin-off from research results being made available to other industrial sectors; the continued contact between the JRC researchers and the national experts; the coordination of LWR safety research with that of the Member States; and, the JRC work on fast breeders to be planned with regard to the R and D programmes of the Fast Reactor European Consortium

Transactions of the nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1991-10-01

This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately

Outline of nuclear power plant project of fast breeder reactor (FBR) 'Monju'

International Nuclear Information System (INIS)

Suzuki, Fumio; Higashide, Noboru

1982-01-01

The first step of the construction of the fast breeder prototype reactor ''Monju'' in Japan was taken by the declaration of agreement of Fukui Prefecture governor in May, 1982, and the succeeding approval by the cabinet council. In this paper, the outline of the project is described. The paper first introduces the development status of FBRs in the world, and next, describes the progress of planning the ''Monju'' project. Initially, ''Monju'' was invited by Shiraki region of Tsuruga City, but the move of village was not agreed. Thus, the center of the site was moved to ''Haseda'' about 1 km north-east. In addition, there were some limitations, and the layout of the facility and the design of public work structures including harbour construction were done under many limiting conditions. However, the site is rather in good conditions, and the design rationalization was attempted as far as possible within a range which does not affect the safety of the nuclear power plant. The outline of the examination related to public works and the siting conditions are reported, dividing them into topographic features, geology, meteorology, sea conditions, and natural environment. As the public works plan, site preparation, roads, harbour installations and water intake and discharge facility are also described. (Wakatsuki, Y.)

Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

International Nuclear Information System (INIS)

Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

1996-01-01

In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

48 CFR 52.232-31 - Invitation To Propose Financing Terms.

Science.gov (United States)

2010-10-01

... Financing Terms. 52.232-31 Section 52.232-31 Federal Acquisition Regulations System FEDERAL ACQUISITION... Clauses 52.232-31 Invitation To Propose Financing Terms. As prescribed in 32.205(b) and 32.206, insert the following provision: Invitation To Propose Financing Terms (OCT 1995) (a) The offeror is invited to propose...

Current activities on improving storage conditions of the research reactor RA spent fuel - Part II

International Nuclear Information System (INIS)

Matausek, M.V.; Kopecni, M.; Vukadin, Z.; Plecas, I.; Pavlovic, R.; Sotic, O.; Marinkovic, N.

1998-01-01

To minimize further corrosion and preserve integrity of aluminum barrels and the stainless steel channel-type containers that were found to contain leaking spent fuel, actions to improve conditions in the existing spent fuel storage pool at the RA research reactor were initiated. Technology was elaborated and equipment was produced and applied for removal of sludge and other debris from the bottom of the pool, filtration of the pool water, sludge conditioning in cement matrix and disposal at the low and medium waste repository at VINCA site. More sophisticated operations are to be performed together with foreign experts. Safety measures and precautions were determined. Subcriticality was proved under normal and/or possible abnormal conditions. (author)

Evaluation of the RSG-GAS Alpha-Beta Aerosol Contaminant Monitor Performance Under Reactor Operation Condition

International Nuclear Information System (INIS)

Hartoyo, Unggul; Setiawanto, Anto; Sumarno, Yulius

2000-01-01

Analysis to evaluate the RSG-GAS alpha-beta aerosol contaminant monitor performance was done. The high potential radiation working area such as in RSG-GAS is important to monitored for personal safety. Further it is necessary to assure that the system monitor is reliable enough under normal conditions as well as emergency condition. The method uses in this analysis are monitoring and comparing with the standard source. The standard course indicator and panel in main control room indicate that the result is 1 x 110 exp 9 Ci/m exp 3. Based on data monitor observation, the RSG-GAS alpha-beta aerosol contaminant monitor system under reactor operation condition has a good enough performance

Reactor power peaking information display

International Nuclear Information System (INIS)

Book, T.L.; Kochendarfer, R.A.

1986-01-01

This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

Energy Technology Data Exchange (ETDEWEB)

Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M., E-mail: frenken@physics.leidenuniv.nl [Huygens-Kamerlingh Onnes Laboratory, Leiden University, P.O. box 9504, 2300 RA Leiden (Netherlands); Ofitserov, A.; Baarle, G. J. C. van [Leiden Probe Microscopy B.V., J.H. Oortweg 21, 2333 CH Leiden (Netherlands)

2014-08-15

To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

Reactor construction programme in Japan

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-04-15

In September last year, the Japanese Government requested the International Atomic Energy Agency to supply three tons of natural uranium for a research reactor, and the Agency has now arranged for its sale to Japan. The metal will be supplied in ingot form and after fabrication it will be used as fuel in a reactor of the natural uranium, heavy water type. The uranium will be obtained from Canada and sold to Japan by IAEA. The Agency had invited tenders for its supply, and after considering the tenders received, the Agency's Board of Governors decided that the Canadian offer to the Agency of three tons of natural uranium free of charge should be accepted and that the selling price to Japan should be US $35. 50 per kilogramme. The price takes into account Article XIV/E of the Agency's Statute which says that the Agency shall establish a scale of charges (including those for storage and handling) for materials furnished to Member States, and that the scale shall be designed to produce revenues to meet expenses in connexion with materials acquired by the Agency and costs of materials and services provided by it under agreements with one or more members. This is the first operation of its kind to be undertaken by the Agency, and the reactor for which the supply is being made will be the first in Japan to be constructed by Japanese scientists and technicians. IAEA's Board of Governors has given the necessary approval to the reactor project for which the Agency is providing assistance

New reactor concepts for new generation of nuclear power plants: an overview, invited paper

International Nuclear Information System (INIS)

Vujic, J.; Greenspan, E.; Milosevic, M.

2006-01-01

The outlook for energy demand underscores the need to increase the share of nuclear energy production. Achieving the vision of sustainable growth of nuclear energy will require development of both advanced nuclear fuel cycles and next generation reactor technologies and advanced reprocessing and fuel treatment technologies. To achieve this vision, the US department of energy (DOE) has adopted new strategy, the Global Nuclear Energy Partnership (GNEP), which integrates earlier programs: the Generation IV Nuclear Energy Systems Initiative (Generation IV), Nuclear Hydrogen Initiative (NHI), and the Advanced Fuel Cycle Initiative (AFCI) with proliferation-resistant spent fuel reprocessing to minimize nuclear waste. Generation IV furthers this vision beyond previous energy systems, such as Generation III+, through incremental improvements in economic competitiveness, sustainability, development of passively safe systems, and breakthrough methods to reduce the routes of nuclear proliferation. This paper summarizes the main characteristics of the six most promising nuclear energy systems identified by the Generation IV Roadmap and reviews some Generation IV system designs for small-side proliferation resistant reactors being developed by University of California at Berkeley. (author)

Two new techniques for the remote evaluation of reactor steel condition - microscopic removal and surface examination

International Nuclear Information System (INIS)

Clayton, R.

Much reactor inspection work involves an assessment of the condition of structural steel. This paper reviews two different techniques which provide information for such an assessment. The first - micro-sample removal (for the measurement of surface oxide thickness and chemical composition) - requires contact with the steel surface, whereas the second - a 'teach and learn' photographic technique (in which a special photogrammatic camera is used to obtain high-quality close-up photographs, to assess surface condition and corrosion growth) can obtain surface information on inaccessible components. (author)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

2012-07-01

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

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.

Nuclear data for reactors. Proceedings of the second international conference. Vol. II

International Nuclear Information System (INIS)

1970-01-01

The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

Nuclear data for reactors. Proceedings of the second international conference. Vol. I

International Nuclear Information System (INIS)

1970-01-01

The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

2014-01-01

The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

International Nuclear Information System (INIS)

Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

1997-01-01

An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

Chernobyl reactor transient simulation study

International Nuclear Information System (INIS)

Gaber, F.A.; El Messiry, A.M.

1988-01-01

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

Characterization of the planktonic microbiome in upflow anaerobic sludge blanket reactors during adaptation of mesophilic methanogenic granules to thermophilic operational conditions

DEFF Research Database (Denmark)

Zhu, Xinyu; Treu, Laura; Kougias, Panagiotis

2017-01-01

Upflow anaerobic sludge blanket (UASB) technology refers to reactor technology where granules, i.e. self-immobilised microbial associations, are the biological catalysts involved in the anaerobic digestion process. During the start-up period, UASB reactors operate at relatively long HRT and there......Upflow anaerobic sludge blanket (UASB) technology refers to reactor technology where granules, i.e. self-immobilised microbial associations, are the biological catalysts involved in the anaerobic digestion process. During the start-up period, UASB reactors operate at relatively long HRT...... and therefore the liquid phase of the reactor becomes a favourable environment for microbial growth. The current study aimed to elucidate the dynamicity of the suspended microbial community in UASB reactors, during the transition from mesophilic to thermophilic conditions. High throughput 16S rRNA amplicon...... sequencing was used to characterize the taxonomic composition of the microbiome. The results showed that the microbial community was mainly composed by hydrolytic and fermentative bacteria. Results revealed relevant shifts in the microbial community composition, which is mainly determined by the operational...

A novel ion transport membrane reactor for fundamental investigations of oxygen permeation and oxy-combustion under reactive flow conditions

KAUST Repository

Kirchen, Patrick

2013-01-01

Ion transport membrane (ITM) reactors present an attractive technology for combined air separation and fuel conversion in applications such as syngas production, oxidative coupling or oxy-combustion, with the promise of lower capital and operating costs, as well higher product selectivities than traditional technologies. The oxygen permeation rate through a given ITM is defined by the membrane temperature and oxygen chemical potential difference across it. Both of these parameters can be strongly influenced by thermochemical reactions occurring in the vicinity of the membrane, though in the literature they are often characterized in terms of the well mixed product stream at the reactor exit. This work presents the development of a novel ITM reactor for the fundamental investigation of the coupling between fuel conversion and oxygen permeation under well defined fluid dynamic and thermodynamic conditions, including provisions for spatially resolved, in-situ investigations. A planar, finite gap stagnation flow reactor with optical and probe access to the reaction zone is used to facilitate in-situ measurements and cross-validation with detailed numerical simulations. Using this novel reactor, baseline measurements are presented to elucidate the impact of the sweep gas fuel (CH4) fraction on the oxygen permeation and fuel conversion. In addition, the difference between well-mixed gas compositions measured at the reactor outlet and those measured in the vicinity of the membrane surface are discussed, demonstrating the unique utility of the reactor. © 2012 The Combustion Institute. Published by Elsevier Inc. All rights reserved.

Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

2013-04-04

Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

Creating intentionally inviting schools through professional development: an appreciative inquiry

Directory of Open Access Journals (Sweden)

G.M. Steyn

2010-07-01

Full Text Available The professional development (PD of teachers plays an important role in schools; it is indispensable for continuous school development. When schools are exposed to new approaches to learning and teaching, teachers are granted the opportunities to change their thinking and behaviour. In 2009, two South African schools with specific inviting characteristics were nominated for the inviting school award given by the International Alliance for Invitational Education (IAIE. However, the inviting characteristics of these schools were not explicitly intentional according to the IE philosophy, therefore they had to follow a professional development programme aimed at raising teachers’ awareness of invitational education (IE. Workshops were held to equip staff members with IE knowledge and skills, and to increase their understanding of their current practices with a view of making them more intentionally inviting. The study focused on the following two questions: What are the positive experiences of teaching staff concerning the current approach to teaching and learning in schools?; and What strategies may be introduced to assist teachers and their schools in becoming intentionally inviting? These two questions are based on appreciative inquiry (AI and IE. A qualitative research design was most appropriate for the purpose of this study. An analysis of the data revealed two categories (the discovery phase: discovering the best of what exists in the school and the dreaming phase: creating a new future on which AI is based.

Corrosion surveillance for research reactor spent nuclear fuel in wet basin storage

International Nuclear Information System (INIS)

Howell, J.P.

1999-01-01

Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the ''Atoms for Peace'' program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on ''Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water'' and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the Receiving Basin for Offsite Fuels (RBOF) at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993

Dynamic analysis of the reactor building for soft (Kozloduy) and hard (Temelin) soil conditions and different seismic loading

International Nuclear Information System (INIS)

Krutzik, N.

1995-01-01

Analyses were conducted for the reactor building to determine the dynamic responses of the coupled system, soil and structure and the forces in the characteristic structural members. This report summarizes the results of structural dynamic analyses derived for soft and hard soil conditions by the modal time history method using substructure models as well as (for soft soil conditions) in the frequency domain using complex (coupled) models of the soil and the structure. The mathematical model of the reactor building is represented as a lumped mass beam model. The capabilities of the soil were represented by means of global frequency independent springs and dampers (substructure models) or by an appropriate final element model. The results of the above-mentioned analysis presented in this report comprise in particular the maximum values of accelerations, displacements and internal forces as well as the acceleration response spectra for the relevant building regions. The time domain (modal time history) calculations were performed for real soil conditions which corresponds to the site Kozloduy (soft) and Temelin (hard). As seismic input data the corresponding free-field data here been used. The dynamic response obtained for the soft-soil conditions using both type of (substructure and complex) models were compared and demonstrated in one plot. Similar comparison were performed for the results obtained for soft and hard soil conditions

Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

Energy Technology Data Exchange (ETDEWEB)

Seifert, H.P.; Ritter, S

2002-02-01

The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

Energy Technology Data Exchange (ETDEWEB)

Block, R.C.; Feiner, F. [American Nuclear Society, La Grange Park, IL (United States)

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately

Cellulase production by Trichoderma harzianum in static and mixed solid-state fermentation reactors under nonaseptic conditions

Energy Technology Data Exchange (ETDEWEB)

Deschamps, F.; Giuliano, C.; Asther, M.; Huet, M.C.; Roussos, S.

1985-09-01

Cellulase production from lignocellulosic materials was studied in solid-state cultivation by both static and mixed techniques under nonaseptic conditions. The effects of fermentation conditions, such as moisture content, pH, temperature, and aeration, on cellulase production by Trichoderma harzianum using a mixture of wheat straw (80%) and bran (20%) were investigated. With a moisture content of 74% and a pH of 5.8, 18 IU filter paper activity and 198 IU endoglucanase activity/g initial substrate content were obtained in 66 hours. The extension from static column cultivation to stirred tank reactor of 65 l capacity gave similar yields of cellulase.

A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

International Nuclear Information System (INIS)

Ramani, D.T.

1977-01-01

The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

Reactor vessel dismantling at the high flux materials testing reactor Petten

International Nuclear Information System (INIS)

Tas, A.; Teunissen, G.

1986-01-01

The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

International Nuclear Information System (INIS)

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

1996-01-01

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

Computation of reactor control rod drop time under accident conditions

International Nuclear Information System (INIS)

Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

1998-01-01

The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

Partial gasification of coal in a fluidized bed reactor: Comparison of a laboratory and pilot scale reactors

Energy Technology Data Exchange (ETDEWEB)

Xiao, R.; Shen, L.H.; Zhang, M.Y.; Jin, B.S.; Xiong, Y.Q.; Duan, Y.F.; Zhong, Z.P.; Zhou, H.C.; Chen, X.P.; Huang, Y.J. [Southeast University, Nanjing (China)

2007-01-15

A 0.1 MWth lab-scale and 2 MWth pilot-scale experimental rigs were constructed to demonstrate the technical feasibility of a new process. The aim of the lab-scale study is to optimize coal partial gasification reactions operating conditions, which were applied in the pilot-scale tests. A comparison between the laboratory and pilot scale experimental results is presented in this paper in order to provide valuable information for scaling-up of the PFB coal partial reactor to industrial applications. The results show that trends and phenomena obtained in the laboratory reactor are confirmed in a pilot plant operating at similar conditions. However, many differences are observed in the two reactors. The higher heat loss in the lab-scale reactor is responsible for higher equivalence ratio (ER) and lower gas heating value at the similar reactor temperature. With respect to the pilot-scale reactor, mass transfer limitation between bubbles and emulsion phase may become important. Hence, longer contact time is required to achieve the same conversions as in the lab-scale reactor. This difference is explained by a significant change of the hydrodynamic conditions due to the formation of larger bubbles.

Regulations for RA reactor operation

International Nuclear Information System (INIS)

1980-09-01

Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

Effect of process operating conditions in the biomass torrefaction: A simulation study using one-dimensional reactor and process model

International Nuclear Information System (INIS)

Park, Chansaem; Zahid, Umer; Lee, Sangho; Han, Chonghun

2015-01-01

Torrefaction reactor model is required for the development of reactor and process design for biomass torrefaction. In this study, a one-dimensional reactor model is developed based on the kinetic model describing volatiles components and solid evolution and the existing thermochemical model considering the heat and mass balance. The developed reactor model used the temperature and flow rate of the recycled gas as the practical manipulated variables instead of the torrefaction temperature. The temperature profiles of the gas and solid phase were generated, depending on the practical thermal conditions, using developed model. Moreover, the effect of each selected operating variables on the parameters of the torrefaction process and the effect of whole operating variables with particular energy yield were analyzed. Through the results of sensitivity analysis, it is shown that the residence time insignificantly influenced the energy yield when the flow rate of recycled gas is low. Moreover, higher temperature of recycled gas with low flow rate and residence time produces the attractive properties, including HHV and grindability, of torrefied biomass when the energy yield is specified. - Highlights: • A one-dimensional reactor model for biomass torrefaction is developed considering the heat and mass balance. • The developed reactor model uses the temperature and flow rate of the recycled gas as the practical manipulated variables. • The effect of operating variables on the parameters of the torrefaction process is analyzed. • The results of sensitivity analysis represent notable discussions which were not done by the previous researches

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

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Design of Mixed Batch Reactor and Column Studies at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Wu, Weimin; Criddle, Craig S.

2015-01-01

We (the Stanford research team) were invited as external collaborators to contribute expertise in environmental engineering and field research at the ORNL IFRC, Oak Ridge, TN, for projects carried out at the Argonne National Laboratory and funded by US DOE. Specifically, we assisted in the design of batch and column reactors using ORNL IFRC materials to ensure the experiments were relevant to field conditions. During the funded research period, we characterized ORNL IFRC groundwater and sediments in batch microcosm and column experiments conducted at ANL, and we communicated with ANL team members through email and conference calls and face-to-face meetings at the annual ERSP PI meeting and national meetings. Microcosm test results demonstrated that U(VI) in sediments was reduced to U(IV) when amended with ethanol. The reduced products were not uraninite but unknown U(IV) complexes associated with Fe. Fe(III) in solid phase was only partially reduced. Due to budget reductions at ANL, Stanford contributions ended in 2011.

Design of Mixed Batch Reactor and Column Studies at Oak Ridge National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Wu, Weimin [Stanford Univ., CA (United States); Criddle, Craig S. [Stanford Univ., CA (United States)

2015-11-16

We (the Stanford research team) were invited as external collaborators to contribute expertise in environmental engineering and field research at the ORNL IFRC, Oak Ridge, TN, for projects carried out at the Argonne National Laboratory and funded by US DOE. Specifically, we assisted in the design of batch and column reactors using ORNL IFRC materials to ensure the experiments were relevant to field conditions. During the funded research period, we characterized ORNL IFRC groundwater and sediments in batch microcosm and column experiments conducted at ANL, and we communicated with ANL team members through email and conference calls and face-to-face meetings at the annual ERSP PI meeting and national meetings. Microcosm test results demonstrated that U(VI) in sediments was reduced to U(IV) when amended with ethanol. The reduced products were not uraninite but unknown U(IV) complexes associated with Fe. Fe(III) in solid phase was only partially reduced. Due to budget reductions at ANL, Stanford contributions ended in 2011.

Natural fission reactors from Gabon. Contribution to the study of the conditions of stability of a natural radioactive wastes storage site (2 Ga)

International Nuclear Information System (INIS)

Pourcelot, L.

1997-01-01

The natural fission reactors of Oklo consists of a core of uraninite (60%) with fission products, embedded in a pure clay matrix. Thus, the aim of geological, mineral, and geochemical studies of the Oklo Reactors is to assess the behaviour of fission products in an artificial waste depository. Previous studies have shown that Reactor Zone 10, located in the Oklo mine, represents an example for an exceptional confinement of fission products since 2 Ga. In reactor Zone 9, located in Oklo open pit, migrations are more important. Reactor ZOne 13 was influenced by a thermal event due to a doleritic intrusion, located some twenty meters far away, one Ga years after fission reaction operations. In this study,we characterized temperature and redox conditions of fluids by using stable isotopes of uraninites and clays. Moreover mineralogical and chemical characteristics were defined. (author)

MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

International Nuclear Information System (INIS)

Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

2011-01-01

In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

HOMOGENEOUS NUCLEAR POWER REACTOR

Science.gov (United States)

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Pressurized water reactors: the EPR project

International Nuclear Information System (INIS)

Py, J.P.; Yvon, M.

2007-01-01

EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

Osiris reactor descriptive report

International Nuclear Information System (INIS)

1976-03-01

OSIRIS is a swimming pool reactor of 70 MW thermal power. Its main purpose is the irradiation of reactor materials in high neutron flux. A description is given of the air conditioning, ventilation, and radioactive gas removal system. (R.L.)

Which life expectancy for EDF reactors?

International Nuclear Information System (INIS)

Carlier, P.; Fluchere, J.; Giraud, B.; Poizat, F.

2008-01-01

After having recalled that the French nuclear reactors have been designed for a 40-year long life with a 30-year depreciation, the authors notice that foreign reactors which are identical or older get a 60 year operation authorization. The authors analyse the age of the nuclear reactors which are currently operating in the world, in terms of age, of country. They also discuss the license renewal process in the USA. In the next part they analyse the conditions of a life extension to 60 years. These conditions concern material monitoring, maintenance strategy for the replacement of components, the condition of installations, personnel management and behaviour. Appendices contain a table of operating reactors which are still being operated and more than 30 year old, the status of license renewal applications to the US NRC

The reactor Cabri

International Nuclear Information System (INIS)

Ailloud, J.; Millot, J.P.

1964-01-01

It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m 3 /h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

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

Influence of geometrical parameters of the VVER-1000 reactor construction elements to internals irradiation conditions

Directory of Open Access Journals (Sweden)

О. M. Pugach

2015-07-01

Full Text Available Investigations to determine the influences of geometrical parameters of the calculational VVER-1000 reactor model to the results of internal irradiation condition determination are carried out. It is shown that the values of appropriate sensitivity matrix elements are not dependent on a height coordinate for any core level, but there is their azimuthal dependence. Maximum possible relative biases of neutron fluence due to inexact knowledge of internal geometrical parameters are obtained for the baffle and the barrel.

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.)

New materials in nuclear fusion reactors

International Nuclear Information System (INIS)

Iwata, Shuichi

1988-01-01

In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

Juergen Kupitz

2002-01-01

This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited 'all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology' (GC(44)/RES/21) and invited Member States to consider to contribute to a task force on innovative nuclear reactors and fuel cycle (GC(44)/RES/22). In response to this invitation, the IAEA initiated an 'International Project on Innovative Nuclear Reactors and Fuel Cycles', INPRO. The Terms of Reference for INPRO were adopted at a preparatory meeting in November 2000, and the project was finally launched by the INPRO Steering Committee in May 2001. At the General Conference in 2001, first progress was reported, and the General Conference adopted a resolution on 'Agency Activities in the Development of Innovative Nuclear Technology' [GC(45)/RES/12, Tab F], giving INPRO a broad basis of support. The resolution recognized the 'unique role that the Agency can play in international collaboration in the nuclear field'. It invited both 'interested Member States to contribute to innovative nuclear technology activities' at the Agency as well as the Agency itself 'to continue it's efforts in these areas'. Additional endorsement came in a UN General Assembly resolution in December 2001 (UN GA 2001, A/RES/56/94), that again emphasized 'the unique role that the Agency can play in developing user requirements and in addressing safeguards, safety and environmental questions for innovative reactors and their fuel cycles' and stressed 'the need for international collaboration in the development of innovative nuclear technology'. As of February 2002, the following countries or entities have become members of INPRO: Argentina

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.

Invited talk on ageing management of boiling water reactors (BWRs)

International Nuclear Information System (INIS)

Prasad, Y.S.R.; Srinivasan, V.S.

1994-01-01

A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs

Invitation and Evaluation of Bids for Nuclear Power Plants

International Nuclear Information System (INIS)

2011-01-01

This publication emphasizes the integrity and interdependence of various activities related to the bid invitation, technical and economic evaluation and contracting, it updates information included in the existing IAEA documents in order to better reflect the developments in the nuclear and energy industry, compiles a more compact and user friendly guidebook integrating the existing IAEA documents on the subject. It provides the necessary information to organize, guide and realize the activities related to the invitation, the technical and economic evaluation of bids, and contracting as an integrated process. Furthermore, this publication indicates how and to what degree the activities preceding the preparation of the bid invitation specification, the evaluation of bids and contracting could influence the process.

77 FR 1618 - Special Conditions: The Boeing Company, Model 767-300; Seats With Inflatable Lapbelts

Science.gov (United States)

2012-01-11

... energy absorption throughout that range of occupants. In addition, the following situations must be... these special conditions effective upon issuance. Comments Invited We invite interested people to take...

Stability criteria and critical runway conditions of propylene glycol manufacture in a continuous stirred tank reactor

Directory of Open Access Journals (Sweden)

Miguel Ángel Gómez

2015-05-01

Full Text Available Here, a new method for the analysis of the steady state and the safety operational conditions of the hydrolysis of propylene oxide with excess of water, in a Continuous Stirred Tank Reactor (CSTR, was developed. For industrial operational typical values, at first, the generated and removed heat balances were examined. Next, the effect of coolant fluid temperature in the critical ignition and extinction temperatures (TCI and TCE, respectively was analyzed. The influence of the heat exchange parameter (hS on coolant and critical temperatures was also studied. Finally, the steady state operation areas were defined. The existence of multiple stable states was recognized when the heat exchange parameter was in the range 6.636 < hS kJ/(min.K < 11.125. Unstable operation area was located between the TCI and TCE values, restricting the reactor operation area to the low stable temperatures.

Australia's replacement research reactor project

International Nuclear Information System (INIS)

Harris, K.J.

1999-01-01

the technical and contractual requirements of the project. A project manager with extensive experience in major defence projects has been contracted and a team of technical and contracts specialists has now completed documentation which has been issued as a Request for Tender to four pre-qualified reactor vendors. The four vendors were selected as a result of a worldwide invitation in mid 1998 for expressions of interest. The tender process is a 'two envelope' system whereby the tenderers are required to submit in one envelope the technical and contractual details of their bid whilst submitting the pricing details in the other envelope. Tenders close on 3 January 2000. A contract is expected to be agreed by mid 2000 with commissioning completed during 2005

Refurbishment programme of the reactor and progress of work

International Nuclear Information System (INIS)

Astruc, J.M.

1992-01-01

the reactor swimming pool. The invitations to tender for the principal equipment and services were issued jointly by ILL and an industrial architect. The expected delays are presented

One piece reactor removal

International Nuclear Information System (INIS)

Chia, Wei-Min; Wang, Song-Feng

1993-01-01

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

Numerical Analysis of The Effect of Hydrodynamics and Operating Conditions on Biodiesel Synthesis in a Rotor-Stator Spinning Disk Reactor

Directory of Open Access Journals (Sweden)

Wen Zhuqing

2017-06-01

Full Text Available A rotor-stator spinning disk reactor for intensified biodiesel synthesis is described and numerically simulated in the present research. The reactor consists of two flat disks, located coaxially and parallel to each other with a gap ranging from 0.1 mm to 0.2 mm between the disks. The upper disk is located on a rotating shaft while the lower disk is stationary. The feed liquids, triglycerides (TG and methanol are injected into the reactor from centres of rotating disk and stationary disk, respectively. Fluid hydrodynamics in the reactor for synthesis of biodiesel from TG and methanol in the presence of a sodium hydroxide catalyst are simulated, using convection-diffusion-reaction multicomponent transport model with the CFD software ANSYS©Fluent v. 13.0. Effect of operating conditions on TG conversion is particularly investigated. Simulation results indicate that there is occurrence of back flow close to the stator at the outlet zone. Small gap size and fast rotational speed generally help to intensify mixing among reagents, and consequently enhance TG conversion. However, increasing rotational speed of spinning disk leads to more backflow, which decreases TG conversion. Large flow rate of TG at inlet is not recommended as well because of the short mean residence time of reactants inside the reactor.

Reactor operation monitor

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1982-01-01

Purpose: To improve the working performance of a reactor by extending the range for the power conditioning due to the control rod operation and flow rate control. Constitution: The results of calculations for the power distribution and the burn-up degree distribution of the reactor core from a reactor performance computer that processes each of measuring signals in a nuclear power plant are used as the inputs for a computing device of the fuel rod power hysteresis to form the power hysteresis for each of the fuel rods up to the present time. The data are used as the inputs for the computing device of the fuel rod performance index, and the fuel rod performance index representing the critical values for the stresses in the fuel rod cladding tubes and the critical values for the duration of the stresses determined from the power hysteresis and the burn-up degree of the fuel rod are calculated for each of the fuel rods. Accordingly, the power conditioning can be carried out upon power-up in the reactor while monitoring the fuel rod performance index f(t) for each of the fuel assemblies, whereby the range for the power conditioning due to the control rod operation and the flow rate control can be extended relative to fuel assemblies in which f(t) is smaller than 1. (Yoshino, Y.)

Safety analysis of Ignalina NPP during shutdown conditions

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2000-01-01

The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

Diagnostic measurement on research reactors

International Nuclear Information System (INIS)

Dach, K.; Zbytovsky, A.

A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

Thermosyphoning in the CANDU reactor

International Nuclear Information System (INIS)

Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

1984-01-01

Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

Pressurised water reactor operation

International Nuclear Information System (INIS)

Birnie, S.; Lamonby, J.K.

1987-01-01

The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

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

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

International Nuclear Information System (INIS)

Grover, S.B.

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

2000-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1998-01-01

Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1996-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

Chemical conditions in the repository for low- and intermediate-level reactor waste

International Nuclear Information System (INIS)

Snellman, M.; Uotila, H.

1984-01-01

The chemical conditions in the proposed repositories for low- and intermediate-level reactor waste at Haestholmen (IVO) and Olkiluoto (TVO) have been discussed with respect to materials introduced into the repository, their possible long-term changes and interaction with groundwater flowing into the repository. The main possible groundwater-rock interactions have been discussed, as well as the role of micro-organisms, organic acids and colloids in the estimation of the barrier integrity. Experimental and theoretical studies have been performed on the basis of the natural groundwater compositions expected at the repository sites. Main emphasis is put on the chemical parameters which might influence the integrity of the different barriers in the repository as well as on the parameters which might effect the release and transport of radionuclides from the repository

Chemistry of fission product iodine under nuclear reactor accident conditions

International Nuclear Information System (INIS)

Malinauskas, A.P.; Bell, J.T.

1986-01-01

The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

Advanced reactor concepts and safety

International Nuclear Information System (INIS)

Lipsett, J.J.

1988-06-01

The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

Conceptual research on reactor core physics for accelerator driven sub-critical reactor

International Nuclear Information System (INIS)

Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

2000-01-01

The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

Science.gov (United States)

Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

2010-07-01

The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

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.

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

Reactor building

International Nuclear Information System (INIS)

Ebata, Sakae.

1990-01-01

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

Pool type liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Guthrie, B.M.

1978-08-01

Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Transmutation of actinides in power reactors.

Science.gov (United States)

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

Gamma Radiation Assessment In Kartini Reactor And Its Vicinity

International Nuclear Information System (INIS)

Yazid, M.; Supriyatni, E.; Maryono; Bastianudin, Aris

2000-01-01

Measurement to calculate dose assesment for gamma radiation in Kartini Reactor and its vicinity has been done whether on operated or un operated condition. Measurement was performed using height pressured ionization chamber, Reuther Stokes RS-112 production. Measurement location was determined based on distance variation inwardly and outwardly of reactor building and its vicinity. The result showed that the average dose rate in the reactor building when un operated is in the range of 11.4-38.6 mu rad/hour and when the reactor operated is 166.4-1910.9 mu rad/hour. While the vicinity of the reactor on operated condition the average dose rate is 34.4-38.6 mu rad/hour in un operated condition is 6.9-7.0 mu rad/hour. This result showed that the reactor operated did not rise the radiation exposure level in its vicinity. From the personnel assesment dose rate of gamma radiation is 28.54 mrem/week on operated condition, 0.90 mrem.week on un operated condition. While dose rate outside the reactor is 0.44 and 0.27 mrem/week for operated and un operated condition consecutively. This dose rate is still below maximum permissible dose than recommended by the national regulation of radiation protection from BAPETEN No. 01/Ka.BAPETEN/V-99

Small power reactor projects in the United States of America and Canada. Information gathered as a result of invitations from Member States

International Nuclear Information System (INIS)

1962-01-01

As part of its activities in connection with the development of nuclear power, and in response to the resolutions adopted by the General Conference, the Agency has been undertaking a continuing study of the technology and economics of small and medium sized power reactors, particularly with reference to the needs of the less-developed countries. This report summarizes the information gathered on the small power reactor projects in the United States of America and Canada, as a result of the opportunity afforded by these Member States to the Agency. It may be recalled that, at the third regular session of the General Conference, the United States Government offered to provide the Agency with relevant technical and economic data on several small power reactor projects of its Atomic Energy Commission. The Agency accepted the offer and since June 1960 it has sent one or two staff members at approximately six-monthly intervals to follow the development of nine power reactor projects in the United States which represent six different reactor systems. Last year, the Agency issued a report summarizing the information obtained through their visits and study of available published literature. The present document, which should be read in conjunction with that document, brings the information up to date and provides additional information on certain phases of the projects already discussed in the last report. Three more power reactor projects are also dealt with, namely the experimental gas-cooled reactor (EGCR), the high temperature gas-cooled reactor (HTGR) and the Hallam nuclear power facility (HNPF). Early in 1962, the Canadian Government expressed its willingness to make available to the Agency relevant information on the NPD and CANDU projects. The coverage of the NPD reactor is based upon the published information supplied by AECL of Canada and the visit by one of the staff members to the NPD site. The Agency wishes to acknowledge with thanks the co-operation extended

Decommissioning of Salaspils nuclear reactor

International Nuclear Information System (INIS)

Abramenkovs, A.; Malnachs, J.; Popelis, A.

2002-01-01

In May 1995, the Latvian Government decided to shut down the Research Reactor Salaspils (SRR) and to dispense with nuclear energy in future. The reactor has been out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH from 1998-1999. he Latvian Government decided on 26 October 1999 to start the direct dismantling to 'green field' in 2001. The results of decommissioning and dismantling performed in 1999-2001 are presented and discussed. The main efforts were devoted to collecting and conditioning 'historical' radioactive waste from different storages outside and inside the reactor hall. All radioactive material more than 20 tons were conditioned in concrete containers for disposal in the radioactive waste depository 'Radons' in the Baldone site. Personal protective and radiation measurement equipment was upgraded significantly. All non-radioactive equipment and material outside the reactor buildings were free-released and dismantled for reuse or conventional disposal. Weakly contaminated material from the reactor hall was collected and removed for free-release measurements. The technology of dismantling of the reactor's systems, i.e. second cooling circuit, zero power reactors and equipment, is discussed in the paper. (author)

14 CFR 11.75 - Does FAA invite public comment on petitions for rulemaking?

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Does FAA invite public comment on petitions... Rulemaking and for Exemption § 11.75 Does FAA invite public comment on petitions for rulemaking? Generally, FAA does not invite public comment on petitions for rulemaking. ...

Scaling of interfacial jump conditions

International Nuclear Information System (INIS)

Quezada G, S.; Vazquez R, A.; Espinosa P, G.

2015-09-01

To model the behavior of a nuclear reactor accurately is needed to have balance models that take into account the different phenomena occurring in the reactor. These balances have to be coupled together through boundary conditions. The boundary conditions have been studied and different treatments have been given to the interface. In this paper is a brief description of some of the interfacial jump conditions that have been proposed in recent years. Also, the scaling of an interfacial jump condition is proposed, for coupling the different materials that are in contact within a nuclear reactor. (Author)

Representation of boundary conditions in thermal reactor global analysis by diffusion theory employing finite difference approximation

International Nuclear Information System (INIS)

Paul, O.P.K.

1978-01-01

An approach to simulate the flux vanishing boundary condition in solving the two group coupled neutron diffusion equations in three dimensions (x, y, z) employed to calculate the flux distribution and keff of the reactor is summarised. This is of particular interest when the flux vanishing boundary in x, y, z directions is not an integral multiple of the mesh spacings in these directions. The method assumes the flux to be negative, hypothetically at the mesh points lying outside the boundary and thus the finite difference formalism for Laplacian operator, taking into account six neighbours of a mesh point in a square mesh arrangement, is expressed in a general form so as to account for the boundary mesh points of the system. This approach has been incorporated in a three dimensional diffusion code similar to TAPPS23 and has been used for IRT-2000 reactor and the results are quite satisfactory. (author)

Invitation strategies and coverage in the population-based cancer screening programmes in the European Union.

Science.gov (United States)

Vale, Diama B; Anttila, Ahti; Ponti, Antonio; Senore, Carlo; Sankaranaryanan, Rengaswamy; Ronco, Guglielmo; Segnan, Nereo; Tomatis, Mariano; Žakelj, Maja P; Elfström, Klara M; Lönnberg, Stefan; Dillner, Joakim; Basu, Partha

2018-03-21

The aim of this study was to describe the compliance of the population-based cancer screening programmes in the European Union Member States to the invitation strategies enumerated in the European Guidelines and the impact of such strategies on the invitational coverage. Experts in screening programme monitoring from the respective countries provided data. Coverage by invitation was calculated as the proportion of individuals in the target age range receiving a screening invitation over the total number of annualized eligible population. The invitation strategies of 30 breasts, 25 cervical and 27 colorectal national or regional population-based screening programmes are described. Individual mail invitations are sent by 28 breasts, 20 cervical and 25 colorectal screening programmes. Faecal occult blood test kits are sent by post in 17 of the colorectal cancer screening programmes. The majority of programmes claimed to have a population registry, although some use health insurance data as the database for sending invitations. At least 95% invitation coverage was reached by 16 breast, six cervical and five colorectal screening programmes. Majority of the programmes comply with the invitation strategies enumerated in the European guidelines, although there is still scope for improvements. Coverage by invitation is below the desirable level in many population-based cancer screening programmes in European Union.

Seismic behaviour of gas cooled reactor components

International Nuclear Information System (INIS)

1990-08-01

On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Nuclear reactor core assembly

International Nuclear Information System (INIS)

Baxi, C.B.

1978-01-01

The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

Simplified numerical simulation of hot channel in sodium cooled reactor

International Nuclear Information System (INIS)

Fonseca, F. de A.S. da; Silva Filho, E.

1988-12-01

The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

Young Women's Experiences of Resisting Invitations to Use Illicit Drugs

Science.gov (United States)

Koehn, Corinne V.; O'Neill, Linda K.

2011-01-01

Ten young women were interviewed regarding their experiences of resisting invitations to use illicit drugs. Hermeneutic phenomenology was used to gather and analyze information. One key theme was the motivations that inspired women to refuse drug offers. Young women resisted drug invitations because of their desires to be authentic, protect their…

Return momentum effect on reactor coolant water level distribution during mid-loop conditions

International Nuclear Information System (INIS)

Seo, Jae Kwang; Yang, Jae Young; Park, Goon Cherl

2001-01-01

An accurate prediction of the Reactor Coolant System( RCS) water level is of importance in the determination of the allowable operating range to ensure safety during mid-loop operations. However, complex hydrualic phenomena induced by the Shutdown Cooling System (SCS) return momentum causes different water levels from those in the loop where the water level indicators are located. This was apparently observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level distribution, a model using a one-dimensional momentum and energy conservation for cylindrical channel, hydraulic jump in operating cold leg, water level build-up at the Reactor Vessel (RV) inlet nozzle, Bernoulli constant in downcomer region, and total water volume conservation has been developed. The model predicts the RCS water levels at various RCS locations during the mid-loop conditions and the calculation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs, in conjuction with the pressure drop throughout the RCS, is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

Reactor core operation management system

International Nuclear Information System (INIS)

Sato, Tomomi.

1992-01-01

Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

Reactor core operation management system

Energy Technology Data Exchange (ETDEWEB)

Sato, Tomomi.

1992-05-28

Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

76 FR 63818 - Special Conditions: Boeing Model 767-400ER Series Airplanes; Seats With Inflatable Lapbelts

Science.gov (United States)

2011-10-14

... inflatable lapbelt must provide a consistent approach to energy absorption throughout that range of occupants... special conditions effective upon issuance. Comments Invited We invite interested people to take part in...

Berger Peter L., Invitation à la sociologie

Directory of Open Access Journals (Sweden)

Nicolas Marquis

2011-03-01

Full Text Available L’invitation à la sociologie de Peter Berger est la réédition d’un ouvrage initiale­ment paru en 1963. Né en 1929, P. Berger est un des grands noms de la sociologie amé­ricaine actuelle, notamment connu chez nous pour un ouvrage coécrit avec Th. Luck­­mann et paru trois ans après la première édition de Invitation à la sociologie : The Social Construction of Reality, qui constitue une étape importante dans l’approche dite du constructivisme social en réactivant la tradition phénoménologique en...

Method for filling a reactor with a catalyst

DEFF Research Database (Denmark)

2013-01-01

The invention relates to a method for filling a reactor with a catalyst for the carbonylation of carbonylated compounds in the gas phase. According to said method, a SILP catalyst is covered with a filling agent which is liquid under normal conditions and is volatile under carbonylation reaction...... conditions, and a thus-treated catalyst is introduced into the reactor and the reactor is sealed....

Neutron noise in nuclear reactors

International Nuclear Information System (INIS)

Blaquiere, A.; Pachowska, R.

1961-06-01

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

The long-term behaviour of cemented research reactor waste under the geological disposal conditions of the Boom Clay Formation: results from leach experiments

International Nuclear Information System (INIS)

Sneyers, A.; Fays, J.; Iseghem, P. van

2001-01-01

The Belgian Nuclear Research Centre SCK-CEN has carried out a number of studies to evaluate the long-term behaviour of cemented research reactor waste under the geological disposal conditions of the Boom Clay Formation. Static leach experiments in synthetic clay water were performed on active samples of cemented research reactor waste. The leach experiments were carried out under anaerobic conditions at two testing temperatures (23 and 85 o C). Leach rates of seven radionuclides ( 60 Co, 90 Sr, 134 Cs, 137 Cs, 144 Ce, 154 Eu and 241 Am) were measured. Most investigated radionuclides are well retained within the cement matrix over a 280 days testing period. Results on the source term of radionuclides were complemented with data on the leaching behaviour of cement matrix constituents as Ca, Si, Al, Na, K, Mg and SO 4 as well as with data from performance assessment calculations and in situ tests. Despite limitations inherent to short-term experiments, combined results from these investigations indicate only limited interactions of disposed research reactor waste with the near field of a geological repository in clay. (author)

Providing thermal-hydraulic boundary conditions to the reactor code TINTE through a Flownex-TINTE coupling - HTR2008-58110

International Nuclear Information System (INIS)

Marais, D.; Greyvenstein, G. P.

2008-01-01

TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time. Copyright ASME 2008. (authors)

Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

DEFF Research Database (Denmark)

Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

2012-01-01

With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...... conditions. Then, the screening of controlled variables and pairing is carried out by an assessment of the effect of the disturbances based on the closed loop disturbance gain plots. Two controlled structures are obtained and benchmarked by their capacity to reject the disturbances before the Anammox reactor....

Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

2017-06-15

Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

University Reactor Matching Grants Program

International Nuclear Information System (INIS)

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-01-01

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

Energy Technology Data Exchange (ETDEWEB)

Nad, Shreya [Department of Electrical and Computer Engineering, Michigan State University, East Lansing, Michigan 48824 (United States); Department of Physics and Astronomy, Michigan State University, East Lansing, Michigan 48824 (United States); Gu, Yajun; Asmussen, Jes [Department of Electrical and Computer Engineering, Michigan State University, East Lansing, Michigan 48824 (United States)

2015-07-15

The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100–260 Torr pressure range and 1.5–2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficiencies (η{sub coup}) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.

Thermohydraulic analysis of pressurized water reactors

International Nuclear Information System (INIS)

Veloso, M.A.

1980-01-01

The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

Reactor condition monitoring and singularity detection via wavelet and use of entropy in Monte Carlo calculation

International Nuclear Information System (INIS)

Kim, Ok Joo

2007-02-01

Wavelet theory was applied to detect the singularity in reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, by wavelet transform after de-noising, singular points can be found easily. To demonstrate this, we generated reactor power signals using a HANARO (a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations and Gaussian noise. We applied wavelet transform decomposition and de-noising procedures to these signals. It was effective to detect the singular events such as sudden reactivity change and abrupt intrinsic property changes. Thus this method could be profitably utilized in a real-time system for automatic event recognition (e.g., reactor condition monitoring). In addition, using the wavelet de-noising concept, variance reduction of Monte Carlo result was tried. To get correct solution in Monte Carlo calculation, small uncertainty is required and it is quite time-consuming on a computer. Instead of long-time calculation in the Monte Carlo code (MCNP), wavelet de-noising can be performed to get small uncertainties. We applied this idea to MCNP results of k eff and fission source. Variance was reduced somewhat while the average value is kept constant. In MCNP criticality calculation, initial guess for the fission distribution is used and it could give contamination to solution. To avoid this situation, sufficient number of initial generations should be discarded, and they are called inactive cycles. Convergence check can give guildeline to determine when we should start the active cycles. Various entropy functions are tried to check the convergence of fission distribution. Some entropy functions reflect the convergence behavior of fission distribution well. Entropy could be a powerful method to determine inactive/active cycles in MCNP calculation

Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

1997-12-31

The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

International Nuclear Information System (INIS)

Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

1997-01-01

The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

1998-12-31

The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

Methodology for the integral comparison of nuclear reactors: selecting a reactor for Mexico; Metodologia para la comparacion integral de reactores nucleares: seleccion de un reactor para Mexico

Energy Technology Data Exchange (ETDEWEB)

Reyes R, R.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: ricarera@yahoo.com.mx

2006-07-01

In this work it was built a methodology to compare nuclear reactors of third generation that can be contemplated for future electric planning in Mexico. This methodology understands the selection of the reactors to evaluate eliminating the reactors that still are not sufficiently mature at this time of the study. A general description of each reactor together with their main ones characteristic is made. It was carried out a study for to select the group of parameters that can serve as evaluation indicators, which are the characteristics of the reactors with specific values for each considered technology, and it was elaborated an evaluation matrix indicators including the reactors in the columns and those indicators in the lines. Since that none reactor is the best in all the indicators were necessary to use a methodology for multi criteria taking decisions that we are presented. It was used the 'Fuzzy Logic' technique, the which is based in those denominated diffuse groups and in a system of diffuse inference based on heuristic rules in the way 'If <condition> Then consequence> ', where the linguistic values of the condition and of the consequence is defined by diffuse groups, it is as well as the rules always they transform a diffuse group into another. Later on they combine all the diffuse outputs to create a single output and an inverse transformation is made that it transfers the output from the diffuse domain to the real one. They were carried out two studies one with the entirety of the indicators and another in which the indicators were classified in three approaches: the first one refers to indicators related with the planning of the plants inside the context of the general electric sector, the second approach includes indicators related with the characteristics of the fuel and the third it considers indicators related with the acting of the plant in safety and environmental impact. This second study allowed us to know the qualities of

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

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

Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

International Nuclear Information System (INIS)

Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

1977-01-01

GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

Maximum neutron flux in thermal reactors

International Nuclear Information System (INIS)

Strugar, P.V.

1968-12-01

Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

Photocatalytic reactors for treating water pollution with solar illumination. I: a simplified analysis for batch reactors

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

Usual applications of photocatalytic reactors for treating wastewater exhibit the difficulty of handling fluids having varying composition and/or concentrations; thus, a detailed kinetic representation may not be possible. When the catalyst activation is obtained employing solar illumination an additional complexity always coexists: solar fluxes are permanently changing with time. For comparing different reacting systems under similar operating conditions and to provide approximate estimations for scaling up purposes, simplified models may be useful. For these approximations the model parameters should be restricted as much as possible to initial physical and boundary conditions such as: initial concentrations (expressed as such or as TOC measurements), flow rate or reactor volume, irradiated reactor area, incident radiation fluxes and a fairly simple experimental observation such as the photonic efficiency. A combination of a new concept: the ''actual observed photonic efficiency'' with ideal reactor models and empirical kinetic rate expressions can be used to provide rather simple working equations that can be efficiently used to describe the performance of practical reactors. In this paper, the method has been developed for the case of a photocatalytic batch reactor (PBR). (orig.)

Invitation

CERN Multimedia

2004-01-01

Monday, 11 October Gala-Concert to celebrate 50 years of CERN Jack Liebeck (violin) and Katya Apekisheva (piano) Works by C. Debussy, S. Prokofiev, and L. van Beethoven CERN, Main Auditorium, 20:00 For CERN staff and their families With the assistance of PPARC (UK) Free tickets available at the CERN Press Office (50-1-048). Maximum four tickets per person. For more information on the performers, see: http://www.jackliebeck.com/CERN50/JL.htm http://www.jackliebeck.com/CERN50/KA.htm The "WIENER SCHUBERTBUND" celebrates CERN's 50th "DEUTSCHE MESSE" By Franz Schubert performed by the "WIENER SCHUBERTBUND" On the occasion of the Organization's Fiftieth Anniversary, Mr Hans-Walter Gérard Schober, Honorary Consul-General of Austria in Geneva, is pleased to invite all those working at CERN to a mass at which Franz Schubert's German Mass will be sung by the Vienna Schubert Choral Society. This renown men's choir, founded in 1863. Eminent composers, including Richard Strauss, Wilhelm Kinzel, a...

Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

Energy Technology Data Exchange (ETDEWEB)

Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

2004-07-01

In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

Corrosion problem in the CRENK Triga Mark II research reactor

International Nuclear Information System (INIS)

Kalenga, M.

1990-01-01

In August 1987, a routine underwater optical inspection of the aluminum tank housing the core of the CRENK Triga Mark II reactor, carried out to update safety condition of the reactor, revealed pitting corrosion attacks on the 8 mm thick aluminum tank bottom. The paper discuss the work carried out by the reactor staff to dismantle the reactor in order to allow a more precise investigation of the corrosion problem, to repair the aluminum tank bottom, and to enhance the reactor overall safety condition

Advanced nuclear reactor and nuclear fusion power generation

International Nuclear Information System (INIS)

2000-04-01

This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

Radiation-hygienic estimation of training reactors location

International Nuclear Information System (INIS)

Konstantinov, Yu.O.; Fedorin, Eh.V.

1978-01-01

The radiation exposure conditions are provided during the normal operation (excluding emergency situations) of four training pool type reactors. Radiation monitoring of the environment near the reactors do not show any increase in external irradiation or in radioactive contamination over what is considered normal radiation background in the locality. Therefore it is possible to judge the potential levels of additional exposure of the population to radiation from the reactors only by means of theoretic modeling of the radiation conditions. Tabular data on maximal levels of this additional radiation are presented, and it is concluded from these data that it is permissible to install training and research reactors up to 3000 kilowatts within large cities, including dwelling areas

Radiation damage effects in solids special topic volume with invited peer reviewed papers only

CERN Document Server

Virk, Hardev Singh

2013-01-01

Public interest and concern about radiation damage effects has increased during recent times. Nuclear radiation proved to be a precursor for the study of radiation damage effects in solids. In general, all types of radiation, e.g. X-ray, gamma ray, heavy ions, fission fragments and neutrons produce damage effects in materials. Radiation damage latent tracks in solids find applications in nuclear and elementary particle physics, chemistry, radiobiology, earth sciences, nuclear engineering, and a host of other areas such as nuclear safeguards, virus counting, ion track filters, uranium exploration and archaeology. Radiation dosimetry and reactor shielding also involve concepts based on radiation damage in solids. This special volume consists of ten Chapters, including Review and Research Papers on various topics in this field.Physical scientists known to be investigating the effects of radiation on material were invited to contribute research and review papers on the areas of their specialty. The topics include...

Nuclear reactor safety system

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1983-01-01

The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

Research and development studies on the seismic behaviour of the PEC fast reactor

Energy Technology Data Exchange (ETDEWEB)

Martelli, Alessandro [ENEA, Fast Reactor Department, Bologna (Italy); [Bologna University, Faculty of Engineering, Nuclear Engineering Section, Bologna (Italy)

1988-07-01

As introduction to the meeting, this paper provides an overview on the extensive research and development studies performed by ENEA, in co-operation with ANSALDO and ISMES, in the framework of the seismic verification of the Italian PEC fast reactor. The purpose is also to stress the reasons why a wide-ranging experimental programme and detailed numerical analysis, validated on the test results, have been performed for the PEC reactor building and the main vessel. Thus, after some notes on the high levels of the design earthquakes adopted for PEC and the important features of fast reactors in general and PEC as a specific case (making it particularly sensitive to seismic excitations), the paper presents the studies performed for the reactor-block, the core and the shutdown system, summarizing their main features and showing some of the main results. Furthermore, the non-negligible feed-backs of the seismic studies on the reactor-block design are recalled, and the needs of checking seismic design analysis of the main vessel and the reactor building are explained. The on-site experimental programme and the related numerical analysis concerning the main vessel and the reactor building are also shortly described: however, specific papers will present more details on these studies, and will also stress the usefulness of the on-site tests performed on the reactor building for the optimization of the PEC seismic monitoring system. Finally, the Italian lecture invited to this meeting will provide an overview on the state-of-the-art on on-site testing and seismic monitoring in Italy, stressing the perspective of adopting methodologies similar to those used for PEC, for nuclear power plants in general. (author)

Nuclear reactor control column

International Nuclear Information System (INIS)

Bachovchin, D.M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

Power oscillations in BWR reactors

International Nuclear Information System (INIS)

Espinosa P, G.

2002-01-01

One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-06-01

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO₂) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

The influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions

International Nuclear Information System (INIS)

Seifert, H.P.; Ritter, S.

2016-01-01

Highlights: • Chloride effects on SCC crack growth in RPV steels under boiling water reactor conditions. • ppb-levels of chloride may result in fast SCC in normal water chemistry environment. • Much higher chloride tolerance for SCC in hydrogen water chemistry environment. • Potential long-term (memory) effects after severe and prolonged temporary chloride transients. - Abstract: The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy reactor pressure vessel steels was evaluated under simulated boiling water reactor conditions. In normal water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity water values within few 100 h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen water chemistry environment is much higher.

Savannah River Site production reactor technical specifications. K Production Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-01

These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

Thermal-hydraulics of wave propagation and pressure distribution under hypothetical steam explosion conditions in the ANS reactor

Energy Technology Data Exchange (ETDEWEB)

Taleyarkhan, R.P.; Georgevich, V.; N-Valenit, S.; Kim, S.H. [Oak Ridge National Lab., TN (United States)

1995-09-01

This paper describes salient aspects of the modeling and analysis framework for evaluation of dynamic loads, wave propagation, and pressure distributions (under hypothetical steam explosion conditions) around key structural boundaries of the Advanced Neutron Source (ANS) reactor core region. A staged approach was followed, using simple thermodynamic models for bounding loads and the CTH code for evaluating realistic estimates in a staged multidimensional framework. Effects of nodalization, melt dispersal into coolant during explosion, single versus multidirectional dissipation, energy level of melt, and rate of energy deposition into coolant were studied. The importance of capturing multidimensional effects that simultaneously account for fluid-structural interactions was demonstrated. As opposed to using bounding loads from thermodynamic evaluations, it was revealed that the ANS reactor system will not be vulnerable to vertically generated missiles that threaten containment if realistic estimates of energetics are used (from CTH calculations for thermally generated steam explosions without significant aluminum ignition).

A new shape reproduction method based on the Cauchy-condition surface for real-time tokamak reactor control

International Nuclear Information System (INIS)

Kurihara, K.

2000-01-01

A new shape reproduction method is investigated on the basis of an applied mathematical approach. An analytically exact solution of Maxwell's equations in a static current field yields an (boundary) integral equation. In application of this equation to tokamak plasma shape reproduction, it is made clear that a Cauchy condition (both Dirichlet and Neumann conditions) on a hypothetical surface is necessarily identified. To calculate the Cauchy condition using magnetic sensor signals, conversion to numerical formulation of this method is conducted. Then, reproduction errors by this method are evaluated through two numerical tests: The first test uses ideal signals produced from a full equilibrium code in the JT-60 geometry, and the second test uses actual sensor signals in JT-60 experiments. In addition, it is shown that positioning and shape of the Cauchy condition surface is insensitive to reproduction error. Finally, this method is clarified to have preferable features for real-time tokamak reactor control

Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

1999-04-01

The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1987-01-01

RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

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

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

2010-01-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

2010-07-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

Impact of second reminder invitation on uptake of screening and cancer detection in BreastCheck.

LENUS (Irish Health Repository)

Fleming, P

2012-01-01

This research aims to quantify the impact of reminder invitations on uptake and cancer detection in Ireland. Examination of BreastCheck\\'s clinical database (2000-2010) to determine number of women screened following first invitation and after reminder; comparison by age group and screening phase with outcomes of recall rate, cancer detection and true positive rates. Of 819,182 first invitations sent 448,974 (54.8%) women attended. 245,157 (66.2%) women attended after reminder invitations, increasing uptake by 29.9% to 694,131 (84.7%) and cancers detected by 1,550 (35%). Women awaiting a reminder were less likely recalled for assessment 9,555 (3.9%) than respondents to first invitation 2,887 (4.04%) (p=0.004). Younger, mainly initial women were more likely recalled for assessment after first invitation. There was no difference between cohorts for cancer detection rate or true positive rate. Reminders increased uptake, supporting international evidence. For programme efficiency attendance at first invitation is optimal. For maximum programme effectiveness attendance must be encouraged with reminders.

Eliciting Information on Sensitive Matters Without Inviting ...

Indian Academy of Sciences (India)

Eliciting Information on Sensitive Matters. Without Inviting Respondents' ... methods based on Randomized Response tech- niques. ... while collecting data on some sensitive issues are well ..... Suppose there is an association of professionals.

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

Electrorefining open-quotes Nclose quotes reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.

1995-01-01

Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined

Study of reactivity of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Rammsy, J.E.M.

1985-01-01

The reactor physics calculations of a 19 module Fluidized Bed Nuclear Reactor using Leopard and Odog codes are performed. The behaviour of the reactor was studied by calculating the reactivity of the reactor as a function of the parameters governing the operational and accidental conditions of the reactor. The effects of temperature, pressure, and vapor generation in the core on the reactivity are calculated. Also the start up behaviour of the reactor is analyzed. For the purpose of the study of a prototype research reactor, the calculations on a one module reactor have been performed. (Author) [pt

A Mechanically Proved and an Incremental Development of the Session Initiation Protocol INVITE Transaction

Directory of Open Access Journals (Sweden)

Rajaa Filali

2014-01-01

Full Text Available The Session Initiation Protocol (SIP is an application layer signaling protocol used to create, manage, and terminate sessions in an IP based network. SIP is considered as a transactional protocol. There are two main SIP transactions, the INVITE transaction and the non-INVITE transaction. The SIP INVITE transaction specification is described in an informal way in Request for Comments (RFC 3261 and modified in RFC 6026. In this paper we focus on the INVITE transaction of SIP, over reliable and unreliable transport mediums, which is used to initiate a session. In order to ensure the correctness of SIP, the INVITE transaction is modeled and verified using event-B method and its Rodin platform. The Event-B refinement concept allows an incremental development by defining the studied system at different levels of abstraction, and Rodin discharges almost all proof obligations at each level. This interaction between modeling and proving reduces the complexity and helps in assuring that the INVITE transaction SIP specification is correct, unambiguous, and easy to understand.

Reactor pressure boundary materials

International Nuclear Information System (INIS)

Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

2002-04-01

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

Calculated and experimental definition of neutron-physical and temperature conditions of material testing in the SM reactor

International Nuclear Information System (INIS)

Toporova, V.G.; Pimenov, V.V.

2004-01-01

Full text: Reactor material science is one of the main scientific directions of the RIAR activities. Particularly, a wide range of materials and products testing under irradiation is performed in reactor facility SM (RF SM). To solve the tasks specified in the technical specification for an experiment, previously, the test conditions are chosen. At the minimum a space-energy distribution of neutrons and heating rate in the materials under test are important as well as temperature conditions of irradiation. The up-to-date software and libraries of nuclear data allow modeling of neutron-material interaction processes to a considerable degree of details and also obtaining a true neutron distribution by calculation methods. As a result of a great scope of work on verification, a calculation model, developed on the basis of a package of applied software MCU (option MCU-4/SM22) and analogue Monte-Carlo method, is widely used at RIAR. The MCU geometric module makes it possible to model the SM core and reflector in three-dimensional geometry with sufficient accuracy and to describe all elements of the channel structure and irradiation device with specimens. The calculation model of RF SM is tested using the results of activation experiments performed in its critical assembly, geometric parameters and structural materials of which correspond completely with the prototype. The difference in the calculated and experimental values is less than 2.5%. Possibilities of the calculated estimation of operating temperature conditions of absorbing elements under irradiation should be considered separately. As the conducted calculations and their analysis show, to define the fuel column temperature correctly, one needs reliable data on thermal-physical parameters of materials, especially ceramic ones, such as titanium, dysprosium or boron carbide. This is very important for boron carbide-absorbing elements for actually all their operation parameters (such as: gas release, swelling

Replacement research reactor for Australia

International Nuclear Information System (INIS)

Miller, Ross

1998-01-01

In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

Development of fault diagnostic technique using reactor noise analysis

International Nuclear Information System (INIS)

Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B.

1999-04-01

The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

Development of fault diagnostic technique using reactor noise analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B

1999-04-01

The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

Fuel deposits, chemistry and CANDU® reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2014-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

Method of operating heavy water moderated reactors

International Nuclear Information System (INIS)

Masuda, Hiroyuki.

1980-01-01

Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

International Nuclear Information System (INIS)

Gan, J.; Stoller, R.E.; Was, G.S.

1998-01-01

A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 approximately 325 C, up to approximately10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

Damage of fuel assembly premature changing in a power reactor

International Nuclear Information System (INIS)

Rudik, A.P.

1987-01-01

Material balance, including energy recovery and nuclear fuel flow rate, under conditions of premature FA extraction from power reactor is considered. It is shown that in cases when before and after FA extraction reactor operates not under optimal conditions damage of FA premature changing is proportional to the first degree of fuel incomplete burning. If normal operating conditions of reactor or its operation after FA changing is optimal, the damage is proportional to the square of fuel incomplete burning

Safety-related parameters for the MAPLE research reactor and a comparison with the IAEA generic 10-MW research reactor

International Nuclear Information System (INIS)

Carlson, P.A.; Lee, A.G.; Smith, H.J.; Ellis, R.J.

1989-07-01

A summary is presented of some of the principle safety-related physics parameters for the MAPLE Research Reactor, and a comparison with the IAEA Generic 10-MW Reactor is given. This provides a means to assess the operating conditions and fuelling requirements for safe operation of the MAPLE Research Reactor under accepted standards

Robust stabilization of burn conditions in subignited fusion reactors using artificial neural networks

International Nuclear Information System (INIS)

Vitela, E. Javier; Martinell, J. Julio

2000-01-01

In this work it is shown that robust burn control in long pulse operations of thermonuclear reactors can be successfully achieved with artificial neural networks. The results reported here correspond to a volume averaged zero-dimensional nonlinear model of a subignited fusion device using the design parameters of the tokamak EDA-ITER group. A Radial Basis Neural Network (RBNN) was trained to provide feedback stabilization at a fixed operating point independently of any particular scaling law that the reactor confinement time may follow. A numerically simulated transient is used to illustrate the stabilization capabilities of the resulting RBNN when the reactor follows an ELMy scaling law corrupted with Gaussian noise. (author)

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

Demand-driven biogas production from sugar beet silage in a novel fixed bed disc reactor under mesophilic and thermophilic conditions.

Science.gov (United States)

Terboven, Christiane; Ramm, Patrice; Herrmann, Christiane

2017-10-01

A newly developed fixed bed disc reactor (FBDR) which combines biofilm formation on biofilm carriers and reactor agitation in one single system was assessed for its applicability to demand-driven biogas production by variable feeding of sugar beet silage. Five different feeding patterns were studied at an organic loading of 4g VS L -1 d -1 under mesophilic and thermophilic conditions. High methane yields of 449-462L N kg VS were reached. Feeding variable punctual loadings caused immediate response with 1.2- to 3.5-fold increase in biogas production rates within 15min. Although variable feeding did not induce process instability, a temporary decrease in pH-value and methane concentration below 40% occurred. Thermophilic temperature was advantageous as it resulted in a more rapid, higher methane production and less pronounced decrease in methane content after feeding. The FBDR was demonstrated to be well-suited for flexible biogas production, but further research and comparison with traditional reactor systems are required. Copyright © 2017 Elsevier Ltd. All rights reserved.

Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

Science.gov (United States)

Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

2012-09-01

The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

Student Theological Research as an Invitation

Directory of Open Access Journals (Sweden)

William Badke

2011-12-01

Full Text Available Seminary students, despite having graduated from bachelors programs, struggle to make sense of the goals, processes, skills, and resources of research as graduate students. Beyond brief introductions to research, a scattered number of seminaries have developed either a separate theological information literacy course or have taken a through-the-curriculum approach to enhancing the information abilities of students. The former, however, separates information literacy from the curriculum, while the latter is difficult to implement and maintain. Living in a world of information glut, seminary professors are finding that traditional information dissemination models of education are becoming less viable. What is more, such models tend to teach students about a discipline rather than inviting them into it. These problems present a unique opportunity to place the teaching of information literacy at the foundation of theological education. With such an approach, students may be invited into the disciplines of their professors and enabled to practice these disciplines, thus becoming equipped to turn knowledge into praxis.

Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

Energy Technology Data Exchange (ETDEWEB)

Fuentes C, P

2003-07-01

This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

Evaluation of the CO2 sequestration capacity for coal fly ash using a flow-through column reactor under ambient conditions

International Nuclear Information System (INIS)

Jo, Ho Young; Ahn, Joon-Hoon; Jo, Hwanju

2012-01-01

Highlights: ► A conceptual in-situ mineral carbonation method using a coal ash pond is proposed. ► CO 2 uptake occurred by carbonation reaction of CO 2 with Ca 2+ ions from coal fly ash. ► The CO 2 sequestration capacity was affected by the solid dosage. ► Seawater can be used as a solvent for mineral carbonation of coal fly ash. - Abstract: An in-situ CO 2 sequestration method using coal ash ponds located in coastal regions is proposed. The CO 2 sequestration capacity of coal fly ash (CFA) by mineral carbonation was evaluated in a flow-through column reactor under various conditions (solid dosage: 100–330 g/L, CO 2 flow rate: 20–80 mL/min, solvent type: deionized (DI) water, 1 M NH 4 Cl solution, and seawater). The CO 2 sequestration tests were conducted on CFA slurries using flow-through column reactors to simulate more realistic flow-through conditions. The CO 2 sequestration capacity increased when the solid dosage was increased, whereas it was affected insignificantly by the CO 2 flow rate. A 1 M NH 4 Cl solution was the most effective solvent, but it was not significantly different from DI water or seawater. The CO 2 sequestration capacity of CFA under the flow-through conditions was approximately 0.019 g CO 2 /g CFA under the test conditions (solid dosage: 333 g/L, CO 2 flow rate: 40 mL/min, and solvent: seawater).

Water reactor fuel behaviour and fission products release in off-normal and accident conditions

International Nuclear Information System (INIS)

1987-09-01

The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

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)

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)

Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

International Nuclear Information System (INIS)

Novelli, A.

1982-01-01

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

The IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO): Status, Ongoing Activities and Outlook

International Nuclear Information System (INIS)

Kupitz, Juergen; Depisch, Frank; Azpitarte, Osvaldo

2004-01-01

The IAEA General Conference (2000) invited 'all interested Member States to combine their efforts under the aegis of the IAEA in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology'. In response to this invitation, the IAEA initiated the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). The overall objectives of INPRO are to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21. century in a sustainable manner, and to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles. In order to fulfil these objectives, the first phase of INPRO dealt with the development of a methodology to assess and compare the performance of innovative nuclear energy systems. This methodology includes the definition of a set of Basic principles, User requirements and Criteria to be met in different areas (Economics, Sustainability and environment, Safety of nuclear installations, Waste management and Proliferation resistance). The result of this phase was presented in a IAEA document (IAEA-TECDOC-1362, Guidance for the evaluation of innovative nuclear reactors and fuel cycles) issued in June 2003. In the present phase of the project, case studies are being carried out in order to validate and improve the developed methodology and the defined set of Basic principles, User requirements and Criteria. This paper shortly summarizes the results published in IAEA-TECDOC-1362 and the ongoing actions related to case studies. Finally, an outlook of INPRO activities is presented. (authors)

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

Nuclear Burning Wave Modular Fast Reactor Concept

International Nuclear Information System (INIS)

Kodochigov, N.G.; Sukharev, Yu.P.

2014-01-01

The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

Online calibration method for condition monitoring of nuclear reactor instrumentations based on electrical signature analysis

International Nuclear Information System (INIS)

Syaiful Bakhri

2013-01-01

Electrical signature analysis currently becomes an alternative in condition monitoring in nuclear power plants not only for stationary components such as sensors, measurement and instrumentation channels, and other components but also for dynamic components such as electric motors, pumps, generator or actuators. In order to guarantee the accuracy, the calibration of monitoring system is a necessary which practically is performed offline, under limited schedules and certain tight procedures. This research aims to introduce online calibration technique for electrical signature condition monitoring in order that the accuracy can be maintained continuously which in turn increases the reactor safety as a whole. The research was performed step by stepin detail from the conventional technique, online calibration using baseline information and online calibration using differential gain adjustment. Online calibration based on differential gain adjustment provides better results than other techniques even tough under extreme gain insertion as well as external disturbances such as supply voltages. (author)

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).

Compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

48 CFR 14.203-2 - Dissemination of information concerning invitations for bids.

Science.gov (United States)

2010-10-01

... 14.203-2 Dissemination of information concerning invitations for bids. Procedures concerning display... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Dissemination of information concerning invitations for bids. 14.203-2 Section 14.203-2 Federal Acquisition Regulations System...

Do invitations for cervical screening provide sufficient information to enable informed choice?

DEFF Research Database (Denmark)

Kolthoff, Sie Karen; Hestbech, Mie Sara; Jørgensen, Karsten Juhl

2016-01-01

or unmentioned. The same applied to other important harms, such as false-positive results and the psychological consequences from an abnormal test result. The majority of invitations took a paternalistic approach. While only two invitations (17%) included a pre-assigned appointment date, eight (70...

Reactor network synthesis for isothermal conditions = Síntese de redes de reatores para condições isotérmicas

Directory of Open Access Journals (Sweden)

Lincoln Kotsuka da Silva

2008-07-01

Full Text Available In the present paper, a computational systematic procedure for isothermal Reactor Network Synthesis (RNS is presented. A superstructure of ideal CSTR and PFR reactors is proposed and the model is formulated as a constrained Nonlinear Programming (NLP problem. Complex reactions (series/parallel reactions are considered. The objective function is based on yield or selectivity, depending on the desired product, subject to different operational conditions. The problem constraints are mass balances in the reactorsand in the considered reactor network superstructure. A systematic computational procedure is proposed and a Genetic Algorithm (GA is developed to obtain the optimal reactor arrangement with the maximum yield or selectivity and minimum reactor volume. Results are as good as or better than those reported in the literature.No presentetrabalho apresenta-se um procedimento computacional para síntese de redes de reatores (SRR operando em condições isotérmicas. Uma superestrutura de rede de reatores formada por reatores ideais CSTR e PFR é proposta e o problema apresenta uma formulação de programação não linear (PNL. São consideradas reações complexas (série/paralelas. A função objetivo é baseada no rendimento ou na seletividade em relação ao produto desejado, sujeito a diferentes condições de operação. As restrições ao problema são provenientes dos balanços de massa e da configuração da superestrutura considerada.No procedimento computacional é proposto um Algoritmo Genético (AG para obtenção do arranjo ótimo de reatores com máximo rendimento ou seletividade com menor volume reacional. Os resultados obtidos são condizentes com os obtidos na literatura.

Decommissioning of the Salaspils Research Reactor

Directory of Open Access Journals (Sweden)

Abramenkovs Andris

2011-01-01

Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

Nuclear power reactor safety

International Nuclear Information System (INIS)

Pon, G.A.

1976-10-01

This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

Reactor on-off antineutrino measurement with KamLAND

NARCIS (Netherlands)

Gando, A.; Gando, Y.; Hanakago, H.; Ikeda, H.; Inoue, K.; Ishidoshiro, K.; Ishikawa, H.; Koga, M.; Matsuda, R.; Matsuda, S.; Mitsui, T.; Motoki, D.; Nakamura, K.; Obata, A.; Oki, A.; Oki, Y.; Otani, M.; Shimizu, I.; Shirai, J.; Suzuki, A.; Takemoto, Y.; Tamae, K.; Ueshima, K.; Watanabe, A.; Xu, B.D.; Yamada, S.; Yamauchi, Y.; Yoshida, H.; Kozlov, A.; Yoshida, S.; Piepke, A.; Banks, T.I.; Fujikawa, B.K.; Han, K.; O'Donnell, T.; Berger, B.E.; Learned, J.G.; Matsuno, S.; Sakai, M.; Efremenko, Y.; Karwowski, H.J.; Markoff, D.M.; Tornow, W.; Detwiler, J.A.; Enomoto, S.; Decowski, M.P.

2013-01-01

The recent long-term shutdown of Japanese nuclear reactors has resulted in a significantly reduced reactor ν¯e flux at KamLAND. This running condition provides a unique opportunity to confirm and constrain backgrounds for the reactor ν¯e oscillation analysis. The data set also has improved

Experimental conditions for determination of the neutrino mass hierarchy with reactor antineutrinos

Directory of Open Access Journals (Sweden)

Myoung Youl Pac

2016-01-01

Full Text Available This article reports the optimized experimental requirements to determine neutrino mass hierarchy using electron antineutrinos (ν¯e generated in a nuclear reactor. The features of the neutrino mass hierarchy can be extracted from the |Δm312| and |Δm322| oscillations by applying the Fourier sine and cosine transforms to the L/E spectrum. To determine the neutrino mass hierarchy above 90% probability, the requirements on the energy resolution as a function of the baseline are studied at sin2⁡2θ13=0.1. If the energy resolution of the neutrino detector is less than 0.04/Eν and the determination probability obtained from Bayes' theorem is above 90%, the detector needs to be located around 48–53 km from the reactor(s to measure the energy spectrum of ν¯e. These results will be helpful for setting up an experiment to determine the neutrino mass hierarchy, which is an important problem in neutrino physics.

An Invitation to Mathematics

CERN Document Server

Schleicher, Dierk

2011-01-01

This "Invitation to Mathematics" consists of 14 contributions, many from the world's leading mathematicians, which introduce the readers to exciting aspects of current mathematical research. The contributions are as varied as the personalities of active mathematicians, but together they show mathematics as a rich and lively field of research. The contributions are written for interested students at the age of transition between high school and university who know high school mathematics and perhaps competition mathematics and who want to find out what current research mathematics is

Meeting of the Technical Working Group on Fast Reactors (TWG-FR) (39th annual meeting). Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

The 39th Annual Meeting of the Technical Working Group on Fast Reactors (TWG FR) was held from 15-19 May 2006 in Beijing, China, at the invitation of the China Institute of Atomic Energy (CIAEA). The meeting was attended by TWG-FR Members and Advisers from the following Member States (MS): Belgium (observer), Brazil, China, France, Germany, India, Italy, Japan, the Republic of Kazakhstan, the Republic of Korea, the Russian Federation, Sweden (observer), the United Kingdom, and the United States. Belarus, Switzerland, the European Commission, and OECD/NEA were unable to participate. Moreover, Prof. Carlo Rubbia, CERN director general emeritus, participated, upon IAEA invitation, in the meeting as distinguished scientist and IAEA expert. Mr. S.C. Chetal, from India (Indira Gandhi Centre for Atomic Research, IGCAR), was appointed chairman. The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 38th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2006 and 2007; - Reviewed the Agency's ongoing information exchange and co-ordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as co-ordination of the TWG-FR's activities with other organizations; - Discuss future joint activities in view of the Agency's Programme and Budget Cycle 2008-2009 (and beyond)

Verification of reactor safety codes

International Nuclear Information System (INIS)

Murley, T.E.

1978-01-01

The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

Energy Technology Data Exchange (ETDEWEB)

Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

2016-12-15

The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

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.)

Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

Energy Technology Data Exchange (ETDEWEB)

Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-06-15

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

Evaluation of six papers from France, F.R. Germany, Italy, Japan, UK, USA, on the need for new safety testing facilities to study fuel-failure under loss-of-flow and transient overpower conditions

International Nuclear Information System (INIS)

Cicognani, C.; Justin, F.; Jakeman, D.; Hannum, W.H.; Kawaguchi, O.; Kessler, G.

1975-01-01

The International Working Group on Fast Reactors invited six countries (France, FRG, Italy, Japan, United Kingdom, USA) to express their views on the importance of fast reactor safety experiments and asked them to give an assessment of the current and future needs of additional safety testing facilities. This papers includes the results of an evaluation of the different views and statements as well as conclusions listed for comparison

Evaluation of six papers from France, F.R. Germany, Italy, Japan, UK, USA, on the need for new safety testing facilities to study fuel-failure under loss-of-flow and transient overpower conditions

Energy Technology Data Exchange (ETDEWEB)

Cicognani, C; Justin, F; Jakeman, D; Hannum, W H; Kawaguchi, O; Kessler, G

1975-08-01

The International Working Group on Fast Reactors invited six countries (France, FRG, Italy, Japan, United Kingdom, USA) to express their views on the importance of fast reactor safety experiments and asked them to give an assessment of the current and future needs of additional safety testing facilities. This papers includes the results of an evaluation of the different views and statements as well as conclusions listed for comparison.

Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

Energy Technology Data Exchange (ETDEWEB)

Ha, Kwang Soon, E-mail: tomo@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Cheung, Fan-Bill [The Pennsylvania State University, University Park, PA 16802 (United States); Park, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Two-phase natural circulation flow induced in insulation gap was investigated. Black-Right-Pointing-Pointer Half-scaled non-heating experiments were performed to evaluate flow behavior. Black-Right-Pointing-Pointer The loop-integrated momentum equation was formulated and solved asymptotically. Black-Right-Pointing-Pointer First-order approximate solution was obtained and agreed with experimental data. - Abstract: The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring. To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

International Nuclear Information System (INIS)

Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

1982-01-01

At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

COUNTER-INTERVENTION, INVITATION, BOTH, OR NEITHER ?

African Journals Online (AJOL)

eliasn

2 See Testimony by David H. Shinn, Somalia: U.S. Govrenemnt Policy and Challenges, ... a state of relative peace—one that can be best described as cold peace. ... denied sending its troops to Somalia until it formally declared war on 24 ...... Lebanese government, the invitation of the USSR by the dubious government of.

Design features of BREST reactors. Experimental work to advance the concept of BREST reactors. Results and plans

International Nuclear Information System (INIS)

Filin, A.I.; Orlov, V.V.; Leonov, V.N.; Sila-Novitskij, A.G.; Smirnov, V.S.; Tsikunov, V.S.

2001-01-01

Principle designs of 300 MW(th) and 1200 MW(th) lead-cooled fast reactors are presented. Reactors of various output are shown to be built using the same principles. In conjunction with increased output and to implement inherent safety concept in BREST-1200 reactor design a number of new solutions, which may be used in BREST-300 concept too, has been taken including: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using Field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by-pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

Nuclear reactor fuel elements

International Nuclear Information System (INIS)

Butterfield, C.E.; Waite, E.

1982-01-01

A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

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

African Journals Online (AJOL)

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

Operational methods of the fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Borges, V.; Sefidvash, F.

1993-01-01

The operational curve of reactivity as a function of porosity of the Fluidized Bed Nuclear Reactor is presented. The strategies for start-up, shut-down and maintaining the reactor critical during operation are described. The inherent safety of the reactor from neutronic point of view under steady state condition is demonstrated. (author)

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

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)

L'invisible invité

OpenAIRE

Frémy, Anne

2012-01-01

Tiré du site Internet de Onestar Press: ""They’re a kind of cabin or hut, very basic and awkwardly assembled, wedged into the space of the balcony between the guardrail and the wall of the building.The vertical sides are made of plywood, canvas or tarp, and are topped by a vegetal roofing that is more or less dry, more or less green, threaded with palm wattle. These huts are opaque and seem to communicate with apartments…" Excerpt from L’Invisible invité by Anne Frémy, english by Bruce Bender...

Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

International Nuclear Information System (INIS)

Fehrenbach, P.J.; Wood, J.C.

1987-06-01

Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

Invitations received from potential predatory publishers and fraudulent conferences: a 12-month early-career researcher experience.

Science.gov (United States)

Mercier, Eric; Tardif, Pier-Alexandre; Moore, Lynne; Le Sage, Natalie; Cameron, Peter A

2018-02-01

This study aims to describe all unsolicited electronic invitations received from potential predatory publishers or fraudulent conferences over a 12-month period following the first publication as a corresponding author of a junior academician. Unsolicited invitations received at an institutional email address and perceived to be sent by predatory publishers or fraudulent conferences were collected. A total of 502 invitations were included of which 177 (35.3%) had subject matter relevant to the recipient's research interests and previous work. Two hundred and thirty-seven were invitations to publish a manuscript. Few disclosed the publication fees (32, 13.5%) but they frequently reported accepting all types of manuscripts (167, 70.5%) or emphasised on a deadline to submit (165, 69.6%). Invitations came from 39 publishers (range 1 to 87 invitations per publisher). Two hundred and ten invitations from a potential fraudulent conference were received. These meetings were held in Europe (97, 46.2%), North America (65, 31.0%), Asia (20.4%) or other continents (5, 2.4%) and came from 18 meeting organisation groups (range 1 to 137 invitations per organisation). Becoming an editorial board member (30), the editor-in-chief (1), a guest editor for journal special issue (6) and write a book chapter (11) were some of the roles offered in the other invitations included while no invitation to review a manuscript was received. Young researchers are commonly exposed to predatory publishers and fraudulent conferences following a single publication as a corresponding author. Academic institutions worldwide need to educate and inform young researchers of this emerging problem. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

Determination of Cl-36 in Irradiated Reactor Graphite

International Nuclear Information System (INIS)

Beer, H.-F.; Schumann, D.; Stowasser, T.; Hartmann, E.; Kramer, A.

2016-01-01

Two of the three research reactors at the Paul Scherrer Institute (PSI), the reactors DIORIT and PROTEUS, contained reactor graphite. Whereas the former research reactor DIORIT has been dismantled completely the PROTEUS is subject to a future decommissioning. In case of the DIORIT the reactor graphite was conditioned applying a procedure developed at PSI. In this case the 36 Cl content had to be determined after the conditioning. The result is reported in this paper. The radionuclide inventory including 36 Cl of the graphite used in PROTEUS was measured and the results are reported in here. It has been proven that the graphite from PROTEUS has a radionuclide inventory near the detection limits. All determined radionuclide activities are far below the Swiss exemptions limits. The graphite from PROTEUS therefore poses no radioactive waste. In contrast, the 36 Cl content of graphite from DIORIT is well above the exemption limits. (author)

Program requirements to determine and relate fuel damage and failure thresholds to anticipated conditions in pressurized water reactors

International Nuclear Information System (INIS)

Loyd, R.F.; Croucher, D.W.

1980-03-01

Anticipated transients, licensing criteria, and damage mechanisms for PWR fuel rods are reviewed. Potential mechanistic fuel rod damage limits for PWRs are discussed. An expermental program to be conducted out-of-pile and in the Engineering Test Reactor (ETR) to generate a safety data base to define mechanistic fuel damage and failure thresholds and to relate these thresholds to the thermal-hydraulic and power conditions in a PWR is proposed. The requirements for performing the tests are outlined. Analytical support requirements are defined

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

International Nuclear Information System (INIS)

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

1998-01-01

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

quantification of emergency action levels for research reactor

International Nuclear Information System (INIS)

Wu Zhongwang; Qu Jingyuan; Liu Yuanzhong; Xi Shuren

2000-01-01

Emergency action level (EAL) technical criteria or parameters for emergency conditions classes. Reference methodology for development of EAL in foreign countries, in process of developed and reviewed emergency plan of home several research reactors, the author thought that should be taken initiating conditions which result in emergency conditions quantified some instrumental readings or alarm thresholds, in order to distinguish and confirm emergency conditions and provide technical bases for emergency response actions. Then based on this principle, revised or developed emergency plans of INET Tsinghua University, promote development of work for emergency plan of research reactors

Dose rate in the reactor room and environment during maintenance in fusion reactors

International Nuclear Information System (INIS)

Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

1995-01-01

According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

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

Investigation of slightly forced buoyant flow in a training reactor

International Nuclear Information System (INIS)

Legradi, G.; Aszodi, A.; Por, G.

2001-01-01

A measurement based on the temperature noise analysis method was carried out in the Training Reactor of the Budapest University of Technology and Economics. The main goals were the estimation of the flow velocity immediately above the reactor core and investigation of the thermal-hydraulical conditions of the reactor, mainly in the core. Subsequently 2D and 3D computations were carried out with the aid of the code CFX- 4.3. The main objective of the 2D calculation was to clarify the thermal-hydraulical conditions of the whole reactor tank with a reasonable computing demand. It was also necessary to accomplish 3D numerical investigations of the reactor core and the space above since three dimensional effects of the flow could only be studied in this way. In addition, obtaining certain boundary conditions of the 3D computations was another significant aim of the 2D investigations. It is important that the results of the noise analysis and the operational measuring system of the reactor gave us a basis for verifying our computations.(author)

Radiation conditions at the training IRT-2000 and IR-100 reactors

International Nuclear Information System (INIS)

Fedorin, Eh.V.; Bronshtejn, I.Eh; Martynov, Yu.N.; Chistyakov, N.I.

1978-01-01

The experience is reviewed of radiation hygiene surveys and radiation safety provision during instructional processes on two training and research nuclear reactors of the IRT-2000 type (No. 1 and No. 2) and on an IR-200 reactor. From an analysis of individual dosimetry data the conclusion is made that the trainees and personnel are exposed mainly to external gamma-radiation and also, to a minor degree, to thermal neutrons and beta-radiation. It has been found that a high level of radiation safety is ensured on the training and research so that research and instruction activities are conducted at annual levels of exposure substantially lower than 0.5 rem in the case of trainees and 5 rem in the case of personnel

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

User's manual for the reactor burnup system, REBUS

International Nuclear Information System (INIS)

Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

1972-01-01

A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

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

Kinetic characterization for hemicellulose hydrolysis of corn stover in a dilute acid cycle spray flow-through reactor at moderate conditions

International Nuclear Information System (INIS)

Jin, Qiang; Zhang, Hongman; Yan, Lishi; Qu, Liang; Huang, He

2011-01-01

The kinetic characterization of hemicellulose hydrolysis of corn stover was investigated using a new reactor of dilute acid cycle spray flow-through (DCF) pretreatment. The primary purpose was to obtain kinetic data for hemicellulose hydrolysis with sulfuric acid concentrations (10-30 kg m -3 ) at relatively low temperatures (90-100 o C). A simplified kinetic model was used to describe its performance at moderate conditions. The results indicate that the rates of xylose formation and degradation are sensitive to flow rate, temperature and acid concentration. Moreover, the kinetic data of hemicellulose hydrolysis fit a first-order reaction model and the experimental data with actual acid concentration after accounting for the neutralization effect of the substrates at different temperatures. Over 90% of the xylose monomer yield and below 5.5% of degradation product (furfural) yield were observed in this reactor. Kinetic constants for hemicellulose hydrolysis models were analyzed by an Arrhenius-type equation, and the activation energy of xylose formation were 111.6 kJ mol -1 , and 95.7 kJ mol -1 for xylose degradation, respectively. -- Highlights: → Investigating a novel pretreatment reactor of dilute acid cycle spray flow-through. → Xylose yield is sensitive to flow rate, temperature and acid concentration. → Obtaining relatively higher xylose monomer yield and lower fermentation inhibitor. → Lumping hemicellulose and xylan oligmers together in the model is a valid way. → The kinetic model as a guide for reactor design, and operation strategy optimization.

Some local dilution transient in a pressurized water reactor

International Nuclear Information System (INIS)

Jacobson, S.

1989-01-01

Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

Scaling the Shear-flow Stabilized Z-pinch to Reactor Conditions

Science.gov (United States)

McLean, H. S.; Schmidt, A.; Shumlak, U.; Nelson, B. A.; Golingo, R. P.; Cleveau, E.

2015-11-01

We present a conceptual design along with scaling calculations for a pulsed fusion reactor based on the shear-flow-stabilized Z-pinch device. Experiments performed on the ZaP device, at the University of Washington, have demonstrated stable operation for durations of 20 usec at ~100kA discharge current for pinches that are ~1 cm in diameter and 100 cm long. The inverse of the pinch diameter and plasma energy density scale strongly with pinch current and calculations show that maintaining stabilization durations of ~7 usec for increased discharge current (~15x) in a shortened pinch (10 cm) results in a pinch diameter of ~200 um and plasma conditions that approach those needed to support significant fusion burn and energy gain (Ti ~ 30keV, density ~ 3e26/m3, ntau ~1.4e20 sec/m3). Compelling features of the concept include operation at modest discharge current (1.5 MA) and voltage (40kV) along with direct adoption of liquid metals for at least one electrode--technological capabilities that have been proven in existing, commercial, pulse power devices such as large ignitrons. LLNL-ABS-674920. This work performed under the auspices of the U.S. Department of Energy ARPAe ALPHA Program by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Thermionic nuclear reactor systems

International Nuclear Information System (INIS)

Kennel, E.B.

1986-01-01

Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

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

Receptivity to sexual invitations from strangers of the opposite gender

DEFF Research Database (Denmark)

Hald, Gert Martin; Hogh-Olesen, Henrik

2010-01-01

This study investigated the primary conclusion from Clark and Hatfield's often cited field experiment ``Consent to Sex with a Stranger'' that men agree to sexual invitations from moderately attractive strangers of the opposite gender more readily than women do. In addition, this study investigated...... whether rates of consent are influenced by a subject's age, relationship status, rating of confederate attractiveness, and type of sexual invitation. A number of moderately attractive confederates of the opposite gender individually approached 173 men and 216 women. After a standard introduction...

Nuclear reactor application for high temperature power industrial processes

International Nuclear Information System (INIS)

Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

1977-01-01

This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

Power limit and quality limit of natural circulation reactor

International Nuclear Information System (INIS)

Zhao Guochang; Ma Changwen

1997-01-01

The circulation characteristics of natural circulation reactor in boiling regime are researched. It is found that, the circulation mass flow rate and the power have a peak value at a mass quality respectively. Therefore, the natural circulation reactor has a power limit under certain technological condition. It can not be increased steadily by continually increasing the mass quality. Corresponding to this, the mass quality of natural circulation reactor has a reasonable limit. The relations between the maximum power and the reactor parameters, such as the resistance coefficient, the working pressure and so on, are analyzed. It is pointed out that the power limit of natural circulation reactor is about 1000 MW at present technological condition. Taking the above result and low quality stability experimental result into account, the authors recommend that the reasonable mass quality of natural circulation reactor working in boiling regime is from 2% to 3% under the researched working pressure

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

Inviting parents to take part in paediatric palliative care research: A mixed-methods examination of selection bias

OpenAIRE

Crocker, Joanna C; Beecham, Emma; Kelly, Paula; Dinsdale, Andrew P; Hemsley, June; Jones, Louise; Bluebond-Langner, Myra

2015-01-01

Background: Recruitment to paediatric palliative care research is challenging, with high rates of non-invitation of eligible families by clinicians. The impact on sample characteristics is unknown. Aim: To investigate, using mixed methods, non-invitation of eligible families and ensuing selection bias in an interview study about parents? experiences of advance care planning (ACP). Design: We examined differences between eligible families invited and not invited to participate by clinicians us...

Utilization of nuclear research reactors

International Nuclear Information System (INIS)

1980-01-01

Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

Initial results from dissolution rate testing of N-Reactor spent fuel over a range of potential geologic repository aqueous conditions

International Nuclear Information System (INIS)

Gray, W.J.; Einziger, R.E.

1998-04-01

Hanford N-Reactor spent nuclear fuel (HSNF) may ultimately be placed in a geologic repository for permanent disposal. To determine whether the engineered barrier system that will be designed for emplacement of light-water-reactor (LWR) spent fuel will also suffice for HSNF, aqueous dissolution rate measurements were conducted on the HSNF. The purpose of these tests was to determine whether HSNF dissolves faster or slower than LWR spent fuel under some limited repository-relevant water chemistry conditions. The tests were conducted using a flowthrough method that allows the dissolution rate of the uranium matrix to be measured without interference by secondary precipitation reactions that would confuse interpretation of the results. Similar tests had been conducted earlier with LWR spent fuel, thereby allowing direct comparisons. Two distinct corrosion modes were observed during the course of these 12 tests. The first, Stage 1, involved no visible corrosion of the test specimen and produced no undissolved corrosion products. The second, Stage 2, resulted in both visible corrosion of the test specimen and left behind undissolved corrosion products. During Stage 1, the rate of dissolution could be readily determined because the dissolved uranium and associated fission products remained in solution where they could be quantitatively analyzed. The measured rates were much faster than has been observed for LWR spent fuel under all conditions tested to date when normalized to the exposed test specimen surface areas. Application of these results to repository conditions, however, requires some comparison of the physical conditions of the different fuels. The surface area of LWR fuel that could potentially be exposed to repository groundwater is estimated to be approximately 100 times greater than HSNF. Therefore, when compared on the basis of mass, which is more relevant to repository conditions, the HSNF and LWR spent fuel dissolve at similar rates

Gender parity trends for invited speakers at four prominent virology conference series.

Science.gov (United States)

Kalejta, Robert F; Palmenberg, Ann C

2017-06-07

Scientific conferences are most beneficial to participants when they showcase significant new experimental developments, accurately summarize the current state of the field, and provide strong opportunities for collaborative networking. A top-notch slate of invited speakers, assembled by conference organizers or committees, is key to achieving these goals. The perceived underrepresentation of female speakers at prominent scientific meetings is currently a popular topic for discussion, but one that often lacks supportive data. We compiled the full rosters of invited speakers over the last 35 years for four prominent international virology conferences, the American Society for Virology Annual Meeting (ASV), the International Herpesvirus Workshop (IHW), the Positive-Strand RNA Virus Symposium (PSR), and the Gordon Research Conference on Viruses & Cells (GRC). The rosters were cross-indexed by unique names, gender, year, and repeat invitations. When plotted as gender-dependent trends over time, all four conferences showed a clear proclivity for male-dominated invited speaker lists. Encouragingly, shifts toward parity are emerging within all units, but at different rates. Not surprisingly, both selection of a larger percentage of first time participants and the presence of a woman on the speaker selection committee correlated with improved parity. Session chair information was also collected for the IHW and GRC. These visible positions also displayed a strong male dominance over time that is eroding slowly. We offer our personal interpretation of these data to aid future organizers achieve improved equity among the limited number of available positions for session moderators and invited speakers. IMPORTANCE Politicians and media members have a tendency to cite anecdotes as conclusions without any supporting data. This happens so frequently now, that a name for it has emerged: fake news. Good science proceeds otherwise. The under representation of women as invited

Propagation calculation for reactor cases

Energy Technology Data Exchange (ETDEWEB)

Yang Yanhua [School of Power and Energy Engineering, Shanghai Jiao Tong Univ., Shanghai (China); Moriyama, K.; Maruyama, Y.; Nakamura, H.; Hashimoto, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-11-01

The propagation of steam explosion for real reactor geometry and conditions are investigated by using the computer code JASMINE-pro. The ex-vessel steam explosion is considered, which is described as follow: during the accident of reactor core meltdown, the molten core melts a hole at the bottom of reactor vessel and causes the higher temperature core fuel being leaked into the water pool below reactor vessel. During the melt-water mixing interaction process, the high temperature melt evaporates the cool water at an extreme high rate and might induce a steam explosion. A steam explosion could experience first the premixing phase and then the propagation explosion phase. For a propagation calculation, we should know the information about the initial fragmentation time, the total melt mass, premixing region size, initial void fraction and distribution of the melt volume fraction, and so on. All the initial conditions used in this calculation are based on analyses from some simple assumptions and the observation from the experiments. The results show that the most important parameter for the initial condition of this phase is the total mass and its initial distribution. This gives the requirement for a premixing calculation. On the other hand, for higher melt volume fraction case, the fragmentation is strong so that the local pressure can exceed over the EOS maximum pressure of the code, which lead to the incorrect calculation or divergence of the calculation. (Suetake, M.)

Modelling of plate-out under gas-cooled reactor (GCR) accident conditions

International Nuclear Information System (INIS)

Taig, A.R.

1981-01-01

The importance of plate-out in mitigating consequences of gas-cooled reactor accidents, and its place in assessing these consequences, are discussed. The data requirements of a plate-out modelling program are discussed, and a brief description is given of parallel work programs on thermal/hydraulic reactor behaviour and fuel modelling, both of which will provide inputs to the plate-out program under development. The representation of a GCR system used in SRD studies is presented, and the equations governing iodine adsorption, desorption and transport round the circuit are derived. The status of SRD's plate-out program is described, and the type of sensitivity studies to be undertaken with the partially-developed computer program in order to identify the most useful lines for future research is discussed. (author)

Fast reactor programme

International Nuclear Information System (INIS)

Plakman, J.C.

1982-01-01

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

Thermionic reactor power conditioner design for nuclear electric propulsion.

Science.gov (United States)

Jacobsen, A. S.; Tasca, D. M.

1971-01-01

Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

Methods and devices prepared to eliminate activation and fission products from PEC reactor cover gas

International Nuclear Information System (INIS)

Caponetti, R.; Gherardi, G.; Petrazzuolo, F.

1987-01-01

The major effort made in Italy for the development of fast nuclear reactor is concentrated in the PEC reactor, whose construction is now in the completion stage. The PEC reactor (Prova Elementi di Combustibile - Fuel Element Testing ) is a sodium-cooled reactor with a power rating of 120 MWt, being built for the purpose of studying the behavior of fuel elements under thermal and neutronic conditions similar to those of fast reactor power stations, whit particular attention to safety aspects. The PEC reactor represents a research instrument particularly suitable for studies and experiments in the following fields: performances of the fuel element and its economical optimization (also with the possibility of testing fuel elements not necessarily based on mixed oxides); experiments in the safety field, not only referred to fuel elements, but also to plant subsystems. The experimental program will cover the research of the limit conditions of the typical parameters, such as cladding temperature, linear power, radiation rate, etc. PEC will also allow researches on new-concept fuel elements and thermal, hydraulic and power transients and cycles foreseen in the commercial power plants under normal, upset and emergency conditions. A number of the solutions regarding the PEC reactor and preparatory approaches to its operation are reported in this paper. In particular the following items are discussed: a description of three cover-gas circuits present in the reactor; an estimate of the contamination conditions foreseen under operating conditions; a description of the equipment for the purification of the cover gas and relative operating conditions. There are three cover-gas circuits present in the PEC reactor. They concern the following sodium circuits: primary reactor, primary emergency reactor and sodium purification primary reactor; secondary reactor, test channel and emergency reactor; primary test channel

Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

Energy Technology Data Exchange (ETDEWEB)

Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

1982-01-01

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

Inviting parents to take part in paediatric palliative care research: a mixed-methods examination of selection bias.

Science.gov (United States)

Crocker, Joanna C; Beecham, Emma; Kelly, Paula; Dinsdale, Andrew P; Hemsley, June; Jones, Louise; Bluebond-Langner, Myra

2015-03-01

Recruitment to paediatric palliative care research is challenging, with high rates of non-invitation of eligible families by clinicians. The impact on sample characteristics is unknown. To investigate, using mixed methods, non-invitation of eligible families and ensuing selection bias in an interview study about parents' experiences of advance care planning (ACP). We examined differences between eligible families invited and not invited to participate by clinicians using (1) field notes of discussions with clinicians during the invitation phase and (2) anonymised information from the service's clinical database. Families were eligible for the ACP study if their child was receiving care from a UK-based tertiary palliative care service (Group A; N = 519) or had died 6-10 months previously having received care from the service (Group B; N = 73). Rates of non-invitation to the ACP study were high. A total of 28 (5.4%) Group A families and 21 (28.8%) Group B families (p research findings. Non-invitation and selection bias should be considered, assessed and reported in palliative care studies. © The Author(s) 2014.

A sustainability analysis of the Brazilian multipurpose reactor project

International Nuclear Information System (INIS)

Obadia, I.J.; Perrotta, J.A.

2010-01-01

The project of a new research reactor in Brazil for radioisotope production, support of the nuclear energy program and scientific research has received a positive sign of the government and is starting to be developed by the Brazilian Commission of Nuclear Energy. International Atomic Energy Agency points out that the implementation of a new research reactor is a major undertaking for a country, requiring an analysis to identify to which extent the conditions of the national nuclear program are proper and adequate to lead to a sustainable research reactor life cycle. This paper introduces the Brazilian Multipurpose Reactor Project (RMB) and describes the sustainability analysis performed, which has shown that the national nuclear infrastructure presents a very favourable condition to the implementation of the RMB project as well as to provide a sustainable life cycle for this new research reactor. (author)

Relative Over Population and the “New Social Question”: an Invitation to Marxian Categories

Directory of Open Access Journals (Sweden)

Ednéia Alves de Oliveira

2010-01-01

Full Text Available This article analyzes the mystified concept that poverty has acquired in the current stage of capitalist accumulation. It relates the growth of overpopulation to a critical perspective in which the conflict between labor and capital is an essential condition for the emergence and permanence of this overpopulation. It looks at the relationship of the general law of capitalist accumulation and at transformations in the order of capital, and then at the naturalized approach that the social question has acquired in this context, inviting a review of Marxian categories.

Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

2014-08-15

Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

RA Research nuclear reactor, Part 1, RA reactor operation and maintenance in 1993, with comparative review for the period 1991 - 1993, Annex 3

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

1993-01-01

RA reactor was not operated during 1993 because of the complete instrumentation exchange. Although it has been planned to exchange the complete instrumentation until the end of 1993, and to start reactor operation in the first half of 1993 this was not fulfilled because the instrumentation was not delivered until the end of 1993. Main activities during past seven years were related to construction of the emergency cooling system; repair and reconstruction of the system for handling the spent fuel and improvement of spent fuel storage conditions; exchange of the aged instrumentation. Other reactor components and systems, reactor core, primary coolant loop and gas circulation system are in good condition concerning future start-up [sr

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

Analysis of SBO accident for a swimming pool reactor

International Nuclear Information System (INIS)

Wang Guimin; Li Weiwei; Li Ning; Guo Wenhui

2015-01-01

The RELAP5/MOD3.3 code was adopted to compute the SBO accident condition of a swimming pool reactor. The coolant flow reversal process was calculated, and the influence of parameters of the flow between the core leakage and components on the flow reversal in the SBO accident condition was analyzed. The calculated results show that in the situation the reactor loses all forced flow, the residual heat of the reactor can be removed by the natural circulation flow, and the fuel subassembly will not be damaged. (authors)

Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

International Nuclear Information System (INIS)

Permana, S.; Takaki, N.; Sekimoto, H.

2007-01-01

Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Thermal embrittlement of reactor vessel steels

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

1995-01-01

As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

A compact reactor/ORC power source

International Nuclear Information System (INIS)

Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

1986-01-01

A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

Thermofluid experiments for Fusion Reactor Safety. Visualization of exchange flows through breaches of a vacuum vessel in a fusion reactor under the LOVA condition

International Nuclear Information System (INIS)

Fujii, Sadao; Shibazaki, Hiroaki; Takase, Kazuyuki; Kunugi, Tomoaki.

1997-01-01

Exchange flow rates through breaches of a vacuum vessel in a fusion reactor under the LOVA (Loss of VAcuum event) conditions were measured quantitatively by using a preliminary LOVA apparatus and exchange flow patterns over the breach were visualized qualitatively by smoke. Velocity distributions in the exchange flows were predicted from the observed flow patterns by using the correlation method in the flow visualization procedures. Mean velocities calculated from the predicted velocity distributions at the outside of the breach were in good agreement with the LOVA experimental results when the exchange flow velocities were low. It was found that the present flow visualization and the image processing system might be an useful procedure to evaluate the exchange flow rates. (author)

Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

International Nuclear Information System (INIS)

1994-08-01

The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

Fuel deposits, chemistry and CANDU reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2013-01-01

'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

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

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

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

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)