International Nuclear Information System (INIS)

Tarapur Atomic Power Station unit-4 is first 540 MWe pressurized heavy water reactor. Heavy water is used as the coolant and moderator. With reactor operation tritium is formed by absorption of neutron by Deuterium atoms. Experience in the 220 MWe indicates that tritium is main contributor to the internal dose of radiation workers. Study on Tritium build up in Primary Heat Transport (PHT) and Moderator (MOD) system was carried out at the initial stage of the operation of the unit-3 and 4. This paper brings out tritium activity buildup in the PHT and MOD systems and its comparison with 220 MWe reactors. This paper helps in estimation of the internal dose contribution to the radiation workers at TAPS 3 and 4. (author)

International Nuclear Information System (INIS)

In Tarapur Atomic Power Station unit-3 and 4, which is 540 MWe pressurized heavy water reactor, tritium is produced in primary heat transport system and moderator system. Tritium is a major contributor to the internal dose. Internal dose contributes about 30% of the collective dose. Internal dose monitoring and its control are important to control the collective dose. Estimation of internal dose is done by analysis of bioassay samples of radiation workers. In a month, about 7000 bioassay samples are analysed for the internal dose assessment during normal operation, and about 12000 during the biennial shut down of the reactor. To enhance the sample preparation and counting performance, minimize the entry errors and reduce the processing time, barcode based label generation system was developed for the internal dose monitoring. This paper discusses about the use of barcode system in the internal dose monitoring at ...

International Nuclear Information System (INIS)

... power reactors pressure tube reactors reactors THERMAL REACTORS.

International Nuclear Information System (INIS)

Tarapur Atomic Power Station Unit-4 is first 540 MWe pressurized heavy water reactor in India. It achieved criticality on 06th March 2005 and then operated at full power i.e 500 MWe. Radiation workers during the normal operation and reactor shutdown are exposed to radiation field. The control of dose rates and the collective dose of the radiation workers is most important for the best performance of the reactor. Experience gained during the operation of the 220 MWe reactors has shown that the Moderator system, primary heat transport system, annulus gas system and moderator cover gas system are the main systems contributing to the dose rate and collective dose. In order to identify the radio nuclides contributing to the radiation field, study was undertaken at TAPS Unit-4. Various samples from the Moderator, primary heat transport system, annulus gas system and ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station 3 and 4 is a 540 MWe Pressurized Heavy Water Reactor. It uses 37 - element natural Uranium dioxide (UO_2) fuel pellets encapsulated inside the cylindrical sheath and are welded to the end plate at each end. During an postulated accident in which part of the fuel bundle are exposed to very high temperature (no means of heat removal) and other are at lower temperature (coolant temperature) possibility of failure of end plate weld due of thermal stresses developed by these relative temperature cannot be ruled out. In this report an attempt is made to study behaviour of fuel bundle under different temperature loading. Modelling of 37 element fuel bundle was done in ANSYS FEM. System was analysed for various sets of temperature loading. The system was analysed for plasticity and creep as material nonlinearity. The total strain, creep strain and stress increase as the temperature increases in upper portion of fuel bundle ...

International Nuclear Information System (INIS)

This system provides data logging facility for axial profile and dynamic profile experiments during low power physics experiments. The data from Cobalt Self Powered Neutron Detectors (SPNDs) are logged via hardware consisting of PC plug-in cards and data of vanadium SPNDs are logged from the Flux Mapping System (FMS) network (multicast data). The data along with programmable details of the experiment is transferred to an Access Database. The system provides online displays and extensive reports from the Access Database as per the requirement from Reactor Physicists. This paper briefly describes the evolvement of requirements (from the experience of operation of TAPS 4 to TAPS 3), software architecture and design. (author)

International Nuclear Information System (INIS)

The Steam Generator (SG) serves as an interface between primary and secondary cycle in Pressurized Heavy Water Reactor (PHWR). Failure of steam generator tubes result in leaking of active heavy water in the secondary closed loop. In Tarapur Atomic Power Station-3 and 4 (TAPS- 3 and 4), Scintillator detectors are provided to detect on line heavy water leakages in SG and moderator heat exchangers by monitoring Nitrogen-16 (16N) and Oxygen-19 (19O) activities. Efficacy of detection of these activities at designed detector position on SG blow down line in presence of background radiation field is analysed theoretically. The count rate of 19O and 16N estimated at the detector position inside Reactor Building (RB) shows that detectors only respond to very high leak rates due to presence of high ambient radiation level even though sensitivity is appreciably good. For detector position in RB in the accessible ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station Unit-3 and 4 (TAPS -3 and 4) are the 540 MWe reactors. Unit-4 attained first criticality on 06th March 2005 and operated for about 230 effective full power days (EFPD). Unit-3 attained first criticality on 21st May 2006 and operated for about 20 EFPD. With the reactor operation radiation field increases on the Primary Heat Transport system equipments, Moderator system equipments and auxiliary system equipments due to deposition of fission products and activation products in different reactor systems. These dose rates significantly contributes to the external exposure and stations collective dose during reactor operation, refueling operation and maintenance activities. A study was undertaken at TAPS 3 and 4 to identify the system equipments showing the significant dose rates and identify the radionuclides present in the primary heat ...

International Nuclear Information System (INIS)

The Flux Mapping System (FMS) of 540 MWe PHWR is a system, which is first of its kind used in Indian PHWRs. It is used to compute a detailed flux/power distribution of the reactor core using modal synthesis method .The paper brings out the high availability features of FMS and the software design philosophy. The paper emphasizes on framework based reusable architectural design, which simplifies and speeds up the development of data acquisition systems. (author)

International Nuclear Information System (INIS)

(1977). United States Parvez, A. Becker, M. Boguslawski, D. Harris,

International Nuclear Information System (INIS)

The device of the present invention improves reactor safety by suppressing lowering of water level in a shroud which surrounds a reactor core, even upon occurrence of rupture of pipelines in an emergency reactor core cooling system in a recycling pump-incorporated type reactor. Namely, an opening of each of cooling systems which forms the emergency reactor core cooling device in a reactor pressure vessel is disposed above the upper end of the reactor core. Further, it also comprises an independent high pressure water injection system, gravitational dropping type water injection system and an automatic depressurization system. With such a constitution, even if rupture of pipelines in the system should be assumed, coolants never flow directly from the shroud which surrounds the ...

International Nuclear Information System (INIS)

Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal ...

Energy Technology Data Exchange (ETDEWEB)

This article gives an account of the recent development of light water reactors new concepts in the world. Different projects are being studied. The CE80+ from Combustion Engineering (CE) is a 1350 MWe-PWR-type reactor whose primary circuit is confined in a spherical metallic containment. This reactor was certified by NRC (national regulatory commission) in mid-1996. The APWR (advanced pressurized water reactor) is developed by MHI (Mitsubishi heavy industries) in a collaboration with Westinghouse, this PWR-type reactor fitted with 4 loops derived from the SP90 model that was developed by Westinghouse during the eighties. 2 units of ABWR (advanced boiling water reactor) were commissioned in Japan in 1996 and 1997, ABWR was certified by NRC in mid-1996. The BWR90+ is developed by ABB-atom (Sweden) and it represents a cautious advanced version ...

International Nuclear Information System (INIS)

Division of Remote Handling and Robotics (DRHR) at Bhabha Atomic Research Centre (BARC) has been working on design and development of Reactivity Control Mechanisms for Nuclear Research Reactors (Dhruva, KAMINI and recently Critical Facility of Advanced Heavy Water Reactor (AHWR)) as well as Power Reactors in India (Pressurized Heavy Water Reactors of 220 MWe at Narora and recently India's first 540 MWe PHWR Unit -1 and 2 at Tarapur). This paper gives a brief account of evolution of reactivity control mechanisms for nuclear research and power reactors in India. (author)

Science.gov (United States)

This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capabi...

International Nuclear Information System (INIS)

This paper consists the radiation streaming analysis of Horizontal Flux Unit (HFU) penetrations in Calandria Vault of 540 MWe PHWR. There are total 7 HFU penetrations on west wall of calandria vault. As these penetrations are in accessible area, a detailed analysis has been carried out to find the neutron and gamma dose rates around these penetrations when reactor is operating. Analysis has been carried out Using the computer code MCNP and DOT-III. Based on the predictions at HFU penetrations, shielding arrangement was recommended. Neutron and gamma dose rate higher than estimated were observed at TAPS-4. This was because of installed shield not being similar to recommended one due to site conditions. Subsequently semi-empirical calculations using measured data were carried out by MCNP to further augment the existing shield taking into consideration the space limitations at site. (author)

Medline Plus

... know if the drainage becomes very clear, like tap water. This may be the first indication that there ...

Energy Technology Data Exchange (ETDEWEB)

This patent describes a nuclear system comprising a containment airspace in which a nuclear reactor pressure vessel is disposed there being a reactor core within the pressure vessel. It comprises a heat exchanger elevated a distance above the pressure vessel; a pool of water surrounding the heat exchanger; means for venting the pool of water to an environment outside the containment; a heat exchanger entry conduit within the containment, the entry conduit having an open lower end communicating with the containment space, and an upper end connected to the heat exchanger, water-containing heated fluid present in the containment airspace incident a pressure vessel loss of coolant event entering and flowing through the entry conduit into the heat exchanger for cooling the fluid to convert water vapor therein to a condensate and separate non-condensable gasses ...

International Nuclear Information System (INIS)

This paper discusses the remote plugging of leaks in inaccessible pipework, with main reference to small leaks which frequently appear in ancillary cooling water circuits of nuclear reactors. Initially developed to cure problems of the pre-stressed concrete pressure vessels of UK reactors, the ZORIC sealant has been used to repair leaking biological shield pipework on six CANDU reactors. ZORIC is based on a water-soluble epoxy resin and an aqueous suspension of a refined mineral clay. This paper describes the evolution of the sealant, the qualification and testing program, and their application to CANDU reactor systems. 2 refs., 6 figs.

Energy Technology Data Exchange (ETDEWEB)

The objective of this research was to determine improved thermal, epithermal, and fast fluxes and several responses at mechanical test surveillance location keys 2, 4, 5, and 7 of the pressure vessel of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) for the beginning of the fuel cycle. The purpose of the research was to provide essential flux data in support of radiation embrittlement studies of the pressure vessel shell and beam tubes at some of the important locations.

Energy Technology Data Exchange (ETDEWEB)

This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

Science.gov (United States)

The results of research on the evaluation of the half-bead weld repair method for use on nuclear reactor components are reviewed from data obtained on thick-section test pieces and intermediate-size pressure vessels. Material properties, the magnitude of residual stresses and the structural behavior of flawed pressure vessels are being obtained to determine the adequacy of the weld repair method for application in thick-section components.

Energy Technology Data Exchange (ETDEWEB)

Comments are made on the Public Inquiry into CEGB's proposal to construct a pressurized water reactor (PWR) at Sizewell, UK. Aspects discussed include: time elapsed and its possible effect on the result; economics of nuclear power plants compared with coal-fired power plants; changes in real sterling/dollar exchange rates; effect of mineworkers' strike; the UK electric power generating system; AGR reactors compared with PWR reactors; extension of Magnox reactor life; radioactive waste management; political decisions.

Energy Technology Data Exchange (ETDEWEB)

This study investigated the effectiveness of dealing with biological waste by composting. In particular, it examined the feasibility of recovering excess thermal energy produced in the process of composting biological waste in terms of mass and energy transport parameters required in the aerated compost bed. An experiment was performed in which a 100 dm{sup 3} adiabatic, leak-tight reactor equipped with a controlled aeration system was constructed to study the temperature and pressure distribution in the bed. Sensors were used to determine the amount and humidity of emitted gases under variable external physical conditions. The perforated bottom of the reactor allowed for bed aeration. As such, the humidity and heat were transported upwards, forced by the air pumped in and by natural convection. In terms of pressure distribution inside the composted and aerated bed, the study results showed that there ...

Energy Technology Data Exchange (ETDEWEB)

The calculated radioactive waste inventories of the Turkey Point pressurized water fission reactor (PWR) and the Starfire conceptual fusion tokamak are compared as a function of time from initial start-up to 10,000 years after decommissioning. Only material out of reactor at least one year is considered. The total activity in Ci/W(th) of the Starfire tokamak is slightly greater than that of the PWR during the active lifetimes of the two reactors and beyond 1000 years. However, using reduced activation materials in Starfire can result in about 1/2000 as much long-lived radioactivity as in the fission reactor. It is stressed that comparison of wastes on this basis is not straightforward, since the radioisotopes and methods required for their disposal are different for fusion and fission reactors. 2 refs., 1 fig., 2 tabs.

Energy Technology Data Exchange (ETDEWEB)

The absence of high-temperature electronics is an obstacle to the development of untapped energy resources (deep oil, gas and geothermal). US natural gas consumption is projected to grow from 22 trillion cubic feet per year (tcf) in 1999 to 34 tcf in 2020. Cumulatively this is 607 tcf of consumption by 2020, while recoverable reserves using current technology are 177 tcf. A significant portion of this shortfall may be met by tapping deep gas reservoirs. Tapping these reservoirs represents a significant technical challenge. At these depths, temperatures and pressures are very high and may require penetrating very hard rock. Logistics of supporting 6.1 km (20,000 ft) drill strings and the drilling processes are complex and expensive. At these depths up to 50% of the total drilling cost may be in the last 10% of the well depth. Thus, as wells go deeper it is increasingly important that drillers are able to monitor conditions ...

International Nuclear Information System (INIS)

The present paper deals with the experimental determination of pressure drop across a four-cusped vertical channel. This geometry represents, ideally, the blockage condition in a typical pressurized water reactor with core degraded by accident. Experiments were performed for both single and two-phase flow. Water was utilized for the single-phase measurements whilst simultaneous flow of air and water simulated the steam-water flow. Observation of the prevailing two-phase flow regime was carried out, so that its mechanism could be fully understood. The averaged void fraction was also measured, by the gamma-ray attenuation technique. A wide range of water and air mass flow rates was covered, so that all flow conditions, possible to exist in a reactor with LOCA, could be investigated. New correlations for pressure drop are proposed. (Author).

International Nuclear Information System (INIS)

SMART (System integrated Modular Advanced ReacTor) , is under development at the Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection system (SIS), and an adoption of 4 trains of passive residual heat removal system (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a steam generator tube rupture (SGTR) is one of the most important initiating events which results in a high core damage frequency. Clear ...

Science.gov (United States)

Liquid Phase Catalytic Exchange (LPCE)- Cryogenic Distillation(CD) process was studied which could be available for an optimal tritium removal process of pressurized heavy water reactor system at Wolsung nuclear power plant in the near future. Based upon ...

International Nuclear Information System (INIS)

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding ...

International Nuclear Information System (INIS)

Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the ...

Energy Technology Data Exchange (ETDEWEB)

An energy saving self-powered industrial dehumidifier for use in a building having a tap water conduit leading from a source of supply to utilization points is described comprising: a dehumidifying apparatus positioned in a stream of air to be dehumidified in the building. The dehumidifying apparatus comprising conduit means for diverting at least a portion of a relatively continuously moving stream of tap water from the tap water conduit through a heat exchanger positioned in a stream of air and returning the diverted portion to the tap water stream upstream of the utilization points. Water vapor in the stream of air will condense onto the heat exchanger when the tap water has a temperature below a dew point temperature of the air stream, thereby dehumidifying the air stream.

Energy Technology Data Exchange (ETDEWEB)

Transient experiments in a Temporal Analysis of Products (TAP) Reactor were performed to elucidate the role of surface oyxgen species in the oxidation of benzene to phenol on ZSM-5 type zeolites with nitrous oxide as a selective oxidant. It was shown by puls experiments with nitrous oxide that the mean lifetime of the generated surface oxygen species is between 0.2s at 500 C and about 4.2 s at 400 C. Afterwards the surface oxygen species desorb as molecular oxygen into the gas phase where total oxidation will take place if hydrocarbons are present. Dual puls experiments consisting of a nitrous oxide puls followed by a benzene puls allowed studying the reactivity of the surface oxygen species formed during the first puls. The observation of the phenol formation was impeded due to the strong sorption of phenol. Multipulse experiments were necessary to reach a pseudo steady state phenol yield. (orig.)

Energy Technology Data Exchange (ETDEWEB)

Private transfers of capital from the Organization of Petroleum Exporting Countries (OPEC) to developing countries are intended to complement private transfers from the Organization of Economic Cooperation and Development (OECD) by tapping the emerging capital markets in the Middle East. Developing countries will be able to diversify their borrowing and gain additional financing. The long-term investment of oil-producing countries will benefit and the pressures on the banking institutions to recycle funds will lessen. Middle East capital markets include international loans and international bonds. The history of the Kuwaiti dinar (KD) bond market, with its advantages for both investors and borrowers, illustrates the successful development of a capital market. Financial intermediation needs to be improved, however, if the Middle East is to become efficient enough to compete with the Euromarkets. Efficiency will require different measures and ...

International Nuclear Information System (INIS)

Transients induced in nuclear power plants from many sources result in one or more fluid conditions changing with time. Fluid conditions of pressure, inlet temperature, inlet flow, or even system power many change separately or in conjunction with each other. The result of the condition change may be one which induces departure from nucleate boiling. An experimental investigation of transient which were intended to achieve Critical Heat Flux was performed at the Heat Transfer Research Facility of Columbia University for Siemens Nuclear Power Corporation. The transients were set up to include broad ranges of flow and pressure conditions near the operating range of pressurized water reactors. Transient events were dominated by varying single conditions and measuring the response of the system and of the rod thermocouples. Because of coupling effects within the test loop, secondary conditions would also ...

Energy Technology Data Exchange (ETDEWEB)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden ...

Energy Technology Data Exchange (ETDEWEB)

The catalytic dehydration of methanol to dimethyl ether was investigated for application on-board a methanol fuelled vehicle. Several catalysts have been tested in a fixed bed reactor. Our approach is to develop a small and efficient reactor converting liquid MeOH under pressure and at low reaction temperatures. (author) 2 figs., 5 refs.

Energy Technology Data Exchange (ETDEWEB)

This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

International Nuclear Information System (INIS)

A non-linear mathematical model of dynamics of horizontal steam generator for nuclear power unit with WWER type reactor is presented. To realize this model the GEMMA-120 simulation language for computer Odra-1204 has been used. Necessity of taking into account disposited thermal storage capacities along tubulation of a primary cycle is demonstrated. A number of lumped elements of reactor division against a required static accuracy of calculations has been determined. (author).

Energy Technology Data Exchange (ETDEWEB)

The integrity of the RPV head and reactor internals was assessed by means of fluid-structural analyses using a coupled method to evaluate the water hammer phenomenon arising from high burnup fuel failure under RIA conditions. The fluid viscosity effect on the water column burst as well as the complex three-dimensional flow paths caused by a core shroud and standpipes were considered in this study. The three analysis scenarios were designed to investigate the above mentioned influential factors separately. In the first scenario, a two-dimensional axisymmetric reactor vessel model without any reactor internals was modeled to assess the influence of the fluid dynamics in the NSC RIA regulatory evaluation. This model has an actual RPV geometry and can be simply separated from other influential factors in order to concentrate only on investigation of the fluid viscosity effect. In the second scenario, a two-dimensional ...

International Nuclear Information System (INIS)

Containment is considered as ultimate safety system and is designed to enclose whole reactor system and prevent the spread of active air-borne fission products. For Pressure and Temperature calculation, Design Basis Accident (Dba) is double ended break of reactor inlet header or main steam line break but activity release studies are done to access its performance following limiting design basis accident i.e. Loss of Coolant Accident (LOCA) and Emergency Core Cooling System (ECCS). In such accident scenario, the core is severely damaged and results in production of steam and hydrogen along with release of activity to containment environment. Containment functions are maintained in such accident, and radiological consequences are within the prescribed limits. (author)

British Library Electronic Table of Contents (United Kingdom)

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25mx0.05m) and 2.59m, respectively, whereas the inclination angle of the riser is 50degree. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels mea...

Energy Technology Data Exchange (ETDEWEB)

This report discusses Test Campaign TC07 of the Kellogg Brown & Root, Inc. (KBR) Transport Reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). The Transport Reactor was operated as a pressurized gasifier during TC07. Prior to TC07, the Transport Reactor was modified to allow operations as an oxygen-blown gasifier. Test Run TC07 was started on December 11, 2001, and the sand circulation tests (TC07A) were completed on December 14, 2001. The coal-feed tests (TC07B-D) were started on January 17, 2002 and completed on April 5, 2002. Due to operational ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The steam generator low water level trip setpoint of KORI NPP units 3 and 4(KNU 3 and 4), three-loop Westinghouse pressurized water reactor, is higher than that of OPR1000. In addition, steam generator downcomer water level in KNU 3 and 4 could fluctuate easily during a transient because of smaller downcomer water inventory, compared to the total water inventory in the steam generator. Due to these reasons, there is a higher possibility of unnecessary reactor trips caused by the steam generator low-low water level in KNU 3 and 4. Its operating history shows that most of reactor trips were caused by steam generator low-low level reactor trip signal. Such reactor trips, especially unnecessary ones, result in time and economic losses. In this paper, a feasibility study was performed to reduce unnecessary reactor trip by changing steam generator ...

Energy Technology Data Exchange (ETDEWEB)

The ability of the /sup 252/Cf-source-driven neutron noise analysis method to measure subcriticality has been demonstrated in a variety of experimental configurations of fissile materials. Calculations for an approximately 4-m-dia configuration of light water reactor (LWR) fuel elements indicated the feasibility of measuring the subcriticality of large, loosely coupled arrays of LWR fuel elements by this same method. These analysis suggested application to the initial loading of both pressurized and boiling water reactors, zero-power testing of reactors (such as shutdown margin measurements after initial loading), light water reactor refueling, and safe storage of LWR spent fuel. In the fuel storage application, direct measurement of subcriticality in the actual fuel storage facilities provides the parameter which is directly related to criticality safety.

British Library Electronic Table of Contents (United Kingdom)

In this work, long-term operation of a pilot scale mixed anaerobic reactor processing crude glycerol and rapeseed meal is discussed. These materials are generated as by-products of biodiesel production. Mixed reactor was operated under mesophilic conditions for the period of 654 days. Total cumulative production of biogas reached 379 m3 (at atmospheric pressure and ambient temperature). Maximum volumetric loading achieved during the operation was 2.17 kg m?3 d?1 for the crude glycerol dose of 2 L. When dosing crude glycerol as a single substrate, average specific production of biogas of 0.76 m3 per L of the g-phase was achieved. The lack of nutrients in the g-phase had to be compensated by an addition of ammonium nitrogen in the form of urea into the reactor. Long term processing of crude ...

International Nuclear Information System (INIS)

Growing public and political interests towards incorporating passive safety features in nuclear installations, let Siempelkamp in late 1987 propose a solution consisting of a prestressed cast-iron pressure vessel and a passive heat removal system, integrated in the reactor cell surrounding the vessel. This solution combines the inherent safety of a prestressed metallic pressure vessel with the advantages of a passive heat removal system and thus constitutes a major step towards the goal of further reducing potential residual risks. The design had to meet the boundary conditions for reactor core and reactor building of the modular 200 MWth pebble bed reactor of Siemens/-KWU. The engineering design showed that many input parameters needed for the finite-element-analysis of the overall structure required a verification by measurements in a well scaled test setup. ...

International Nuclear Information System (INIS)

For the layout of a pressure suppression system of a light water cooled reactor (boiling water reactor) it is important to know the time dependent behavior of the vent clearing transient after a loss-of-coolant accident for two main reasons: time of the end of the vent clearing transient influences strongly the pressure and temperature maxima in the drywell and wetwell. Time-dependent behavior of the vent clearing transient influences pressure loads in the condensation pool of the wetwell and therefore pressure induced stresses to the structure. The time-dependent behavior of the water masses in the vent pipes and wetwell are described by the basic equations for a nonstationary incompressible friction flow: momentum equation, continuity equation and a correlation for the variation of the state of the gas volume in the wetwell above the water level. After many ...

Science.gov (United States)

... increasing its economic results, tapping mew financial sources and augmenting its income, but also a basic task in financial and tax management. ...

Energy Technology Data Exchange (ETDEWEB)

Small circular and noncircular channels are representative of flow passages in compact evaporators and condensers. This paper describes results of an experimental study on heat transfer to the flow boiling of refrigerant- 12 in a small circular tube of diameter = 2.46 mm. The objective of the study was to assess the effect of channel size on the heat transfer coefficient and to obtain additional insights relative to the heat transfer mechanisms. The flow channel was made of brass and had an overall length of 0.9 m. The channel wall was electrically heated, and temperatures were measured on the channel wall and in the bulk fluid stream. Voltage taps were located at the same axial locations as the stream thermocouples to allow testing over an exit quality range of 0.21 to 0.94 and a large range of mass flux (63 to 832 kg/m{sup 2}s) and heat flux (2.5 to 59 kW/m{sup 2}). Saturation pressure was nearly constant, averaging 0.82 MPa for most of the ...

International Nuclear Information System (INIS)

This report describes a neutron fluence assessment performed for the Kori Unit 3 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. After Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 3 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of ...

Energy Technology Data Exchange (ETDEWEB)

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water - 310degC and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been ...

Energy Technology Data Exchange (ETDEWEB)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation ...

Science.gov (United States)

The Simplified Boiling Water Reactor (SBWR) designed by General Electric is a natural circulation reactor with enhanced safety features for potential accidents. It has a strong coupling between power and flow in the reactor core, hence the neutronic coupling with thermal-hydraulics is specially important. The potential geysering instability during the early part of a SBWR startup at low flow, low power and low pressure is of particular concern. The RAMONA-4B computer code developed at Brookhaven National Laboratory (BNL) for the SBWR has been used to simulate a SBWR startup transient and evaluate its stability, using a simplified four-channel representation of the reactor core for the thermal-hydraulics. This transient was run for 20,000 sec (5.56 hrs) in order to cover the essential aspect of the SBWR startup. The simulation showed that the SBWR startup was a very challenging event ...

International Nuclear Information System (INIS)

Inherent boron dilution can occur in case of a Small Break LOCA when low borated water is mainly accumulated in the U-legs due to reflux boiling in the Steam Generator tubes after the loss of natural circulation. The restart of the natural circulation may lead to criticality because of the injection of these low borated slugs towards the core. To evaluate this potential risk, the boron concentration at the core inlet has to be known which makes necessary to estimate the mixing phenomena in the cold leg, in the downcomer and in the lower plenum: CFD calculations are required. First of all the validation of CFX5 CFD code on the relevant phenomena of inherent boron dilution has been established (UPTF TRAM C3 test). Then, an application to the 900 MW French Pressurized Water Reactor series has been performed. (authors)

International Nuclear Information System (INIS)

In a 540 MWe PHWR reactor at TAPP-4 the pressuriser has been incorporated in the PHT pressure control system to provide the necessary vapour cushion for PHT main circuit to reduce pressure variations due to transients involving swell and shrinkage. Need for the Pressuriser is due to the large inventory in PHT main circuit and relatively large heat source. The incorporation of Pressuriser is one of the modifications while upgrading from prevalent operating 220 MWe reactors to the present 540 MWe at TAPP-4. The sizing, design and selection of the 540 MWe Pressuriser has been done w.r.t swell/shrinkage requirement during normal/transient operation of the reactor. In this paper the commissioning and operating experience of the Pressuriser in conjunction with the PHT pressure control system is presented. During light water commissioning of PHT circuit some major ...

International Nuclear Information System (INIS)

Generally, Hardening and irradiated brominating phenomena are occurred in the reactor vessel under operation conditions by atomic cavities and creation of impurity atoms which are led by high fast neutron flux. To assure the mechanical integrity of pressure vessel until the end of power plant life after monitoring the sample specimens on the vessel inside, a series of tests is performed over the retrieved surveillance capsule to examine the changes according to the plant operation in accordance with regulations. Monitoring surveillance capsules attached to neutron shield wall of outer core are consists of impact sample, tensile sample and temperature monitor

International Nuclear Information System (INIS)

Several design options for improved passive circulation flow have been investigated for use in small, modular liquid metal cooled reactors (LMRs). The purpose is to enhance the transition to natural convection cooling following loss of forced circulation flow, reducing thermal transients experienced by the fuel and possibly eliminating the need for emergency pony-motor flow. Design details to minimize pressure drops may also enhance maximum equilibrium power levels possible under natural circulation only.

International Nuclear Information System (INIS)

A vortex diode has been developed as a reverse flow limiter in the primary circuit of an advanced gas cooled reactor. In addition to the development work on a prototype diode to optimise performance and geometry, measurements were also made on an available experimental diode of similar size with pressure differences up to 4 MPa and temperatures up to 600 K using nitrogen, argon and carbon dioxide as the test fluids. Correlation of data from all tests was satisfactorily obtained using isentropic one-dimensional nozzle flow equations. (author).

International Nuclear Information System (INIS)

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated ...

Energy Technology Data Exchange (ETDEWEB)

In relation to coal gasification and combustion under high pressure as highly efficient coal utilization, the effect of total pressure and sintering on the SO2 capture ability of Ca-ion exchanged coal and other desulfurizing agents were studied. In experiment, specimens were filled into a small pressurized reactor to heat them under high-pressure N2 atmosphere. After the completion of combustion reaction of char at 850{degree}C, SO2, CO2 and CO gases were measured at an outlet while flowing SO2/N2. As the experimental result, all of the S content in Ca-ion exchanged coal was not absorbed by Ca content in coal during pyrolysis and combustion, resulting in discharge of 36% of the S content. Since Ca-ion exchanged coal is fast in combustion reaction, most of the S content was desulfurized by coal ash. The ash content yielded from Ca-ion exchanged coal was more excellent in SO2 capture ...

International Nuclear Information System (INIS)

Personal monitoring and dose assessment of all radiation workers is an essential regulatory requirement as per radiation safety procedures of AERB and operating stations. The occupational workers of TAPS 1 and 2 and TAPS 3 and 4 are monitored for internal contamination due to high energy gamma emitters by whole body counting

Energy Technology Data Exchange (ETDEWEB)

Sasol, a South-African company, has converted coal into synthetic fuels for 40 years; in order to increase by 6 pc its annual production (40 Mt coal are processed), Sasol has launched the Sky High project and intends to replace the ancient Sasol Synthol reactors by Advanced Synthol reactors using conventional fluidized beds instead of circulating fluidized beds. Coal is first gasified under high pressure and high temperature conditions (the Secunda site uses the Fischer-Tropsch conversion process) and is then processed in the reactor through a single phase process using an iron-base catalyst, leading to significant cost reductions.

Energy Technology Data Exchange (ETDEWEB)

A cold-leg small-break loss-of-coolant accident (SBLOCA) experiment was performed on the PACTEL facility to study the behavior of natural circulation in a VVER-440 reactor geometry. The facility is a volumetrically scaled (1:305) integral test loop simulating the VVER-440 reactors used in Finland. The test results were used to assess the computer codes RELAP5/MOD3.1 and APROS 3.0 for VVER reactors. The behavior of the horizontal steam generator and the effect of the hot-leg loop seal were of particular interest. The specific parameters to be compared included the primary pressure and the downcomer mass flow rate.

Energy Technology Data Exchange (ETDEWEB)

The combustin of chloroethane is modeled as a stirred reactor so that we can study critical emission characteristics of the reactor as a function of residence time. We examine important operating conditions such as pressure, temperature, and equivalence ratio and their influence on destructive efficiency of chloroethane and production of other chlorinated products. The model uses a detailed chemical kinetic mechanism that we have developed previously for C{sub 3} hydrocarbons. We have added to this mechanism the chemical kinetic mechanism for C{sub 2} chlorinated hydrocarbons developed by Senkan and coworkers. Some reactions have been added to Senkan's mechanism and some of the reaction-rate expressions have been updated to reflect recent developments in the literature. In the modeling calculations, sensitivity coefficients are determined to find which reaction-rate constants have the largest effect on destructive ...

Energy Technology Data Exchange (ETDEWEB)

The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

International Nuclear Information System (INIS)

Mitsubishi is developing a new type of accumulator incorporating the technology of fluidics as one of the seeds for the improved safety of the newly constructed pressurized water reactor plants. This accumulator employs a vortex flow control device, called a vortex damper, as a fluidic device to simplify the safety systems. A fundamental experimental study with a one-fifth scale model and confirmation tests with a one-third scale model to develop the vortex damper have been carried out, and satisfactory results have been achieved. The results of the confirmation tests under the prototype pressure conditions agree well with the basic tests. The flow rate ratio can be 5 to 6. The pressure loss coefficient in the large flow rate period is 8. A cavitation factor is the main parameter of the flow rate coefficient.

Energy Technology Data Exchange (ETDEWEB)

The theoretical and experimental investigation of the mechanism of SO2 and H2S removal by CaCO3 -based sorbents (limestones and dolomites) in pressurized uidized-bed coal combustors (PFBC) and high pressure gasiers, respectively, is the main objective of this study. It is planned to carry out reactivity evolution experiments under simulated high pressure conditions or in high pressure thermogravimetric and, if needed, uidized- bed reactor (high pressure) arrangements. The pore structure of fresh, heat-treated, and half-calcined solids (dolomites) will be analyzed using a variety of methods. Our work will focus on limestones and dolomites whose reaction with SO2 or H2S under atmospheric conditions has been studied by us or other research groups in past studies. Several theoret- ical tools will be employed to analyze the obtained experimental data including a ...

International Nuclear Information System (INIS)

The formation and growth of silicon-nanoparticles from silane in a microwave reactor was investigated. Experiments were performed for the following conditions: precursor concentration 380-2530 ppm, pressures of 20-30 mbar, microwave powers 120-300 W. The formed particles were examined in-situ with a particle mass spectrometer. Additionally, particles were collected on grids and analyzed by transmission electron microscopy, X-ray diffraction, and by determining the specific surface area by BET. The particle size was found to be in the range of 5-8 nm in diameter. A simple model was used to simulate the particle formation processes taking place inside the reactor. The microwave energy coupled into the reactor flow was treated as a spatially distributed energy source resulting in a local temperature increase. The particles were assumed to have a monodisperse size distribution. To allow an approximation of ...

International Nuclear Information System (INIS)

Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

International Nuclear Information System (INIS)

The radiation life time of reactor pressure vessel (RPV) is the most important limiting factor for the term of exploitation of the whole power unit. The main degradation mechanism of RPV metal is the neutron induced embrittlement. Processes of radiation ageing running in RPV metal lead to fracture toughness decrease and to increased probability of brittle fracture of the vessel under thermal shocks. This explains the importance of RPV integrity assessment and rest life time management

International Nuclear Information System (INIS)

Some of the origins of corrosion encountered in the secondary side of pressurized water reactor steam generators are:-sludge accumulation (a mixture of metal oxides, primarily magnetite and copper) on tube sheet and attack of tube support plates by aggressive impurities leading to denting. Although Electricite de France has not suffered from these problems, it has developed a chemical cleaning process to dissolve corrosion products at both locations. (author).

International Nuclear Information System (INIS)

For the first time in China, Daya Bay Nuclear Power Station applied the advanced technology of worldwide commercial pressurized reactors to the in-core detectors, the leading excore six-chamber instrumentation for precise axial power distribution, and the related data processing. Described in this article are the neutron flux measurement in Daya Bay Nuclear Power Station, and the detailed data processing.

International Nuclear Information System (INIS)

Gadolinium nitrate has been employed in Indian nuclear reactors for the first time as soluble neutron poison in the heavy water moderators of the 540 MWe PHWRs TAPS 3 and 4, as a fast acting secondary shut down system (SDS-2); and also for reactivity shim. For this purpose, the moderator purification system is currently equipped with special ion-exchange columns/schemes, developed by present authors. However, for gadolinium removal from moderator in the post SDS-2 scenario, the two stage ion-exchange - cation bed operation followed by mixed bed operation - results in low pH conditions persisting in the moderator for a few hours, which gives rise to certain operational problems. The present paper describes a mixed bed ion-exchange scheme employing macro-porous strong acid cation and macro-porous weak base anion resins, which has been developed to eliminate acidic conditions and gives a better pH control. The cation to anion capacity ratio in the ...

International Nuclear Information System (INIS)

This document is a synthesis of data and analysis concerning natural circulation cooling in US Pressurized Water Reactors during off-normal operation and accident transients. Its objective is the integration of important research findings concerning PWR natural circulation phenomena into a single reference document. Sources of information include the Nuclear Regulatory Commission, reactor vendors, utility sponsored research groups, utilities, national laboratories, research reports, meeting papers, archival literature, and foreign sources. Three modes of natural circulation are discussed: single-phase, two-phase, and reflux/boiling condensation. General characteristics, analytical expressions, noncondensible gas effects, secondary effects, and nonuniform flow are described with regard to each of the natural circulation modes. Plant operational data, tests in scaled experimental facilities, and analysis with thermal ...

International Nuclear Information System (INIS)

The Gross Decontamination Experiment was conducted on various levels and surfaces of the TMI-2 Reactor Building during February and March 1982 and was designed to investigate the effectiveness of various surface decontamination techniques. The polar crane, D-rings, missile shields, refueling canal, fueling bridge, major equipment, floors and some walls were flushed with low pressure water. Water lances were directed manually and applied water at temperatures between ambient and 60"0C at a flow rate of about 95 liters per minute. In addition, floor surfaces on the 305-ft elevation and floor surfaces and major equipment on the 347-ft elevation were sprayed with high pressure water (floors in the Reactor Building are designated by their elevations above sea level). The water pressure in this case varied between 13.8 and 41.4 mPa and water temperature was at a maximum 60"0C. Certain ...

Energy Technology Data Exchange (ETDEWEB)

CeO{sub 2}-TiO{sub 2} catalysts are prepared by coprecipitation method, and the activity and stability in the catalytic wet air oxidation (CWAO) of phenol are investigated in a batch reactor and packed-bed reactor. CeO{sub 2}-TiO{sub 2} mixed oxides show the higher activity than pure CeO{sub 2} and TiO{sub 2}, and CeO{sub 2}-TiO{sub 2} 1/1 catalyst displays the highest activity in the CWAO of phenol. In a batch reactor, COD and TOC removals are about 100% and 77% after 120 min in the CWAO of phenol over CeO{sub 2}-TiO{sub 2} 1/1 catalyst at reaction temperature of 150 {sup o}C, the total pressure of 3 MPa, phenol concentration of 1000 mg/L, and catalyst dosage of 4 g/L. In a packed-bed reactor using CeO{sub 2}-TiO{sub 2} 1/1 particle catalyst, over 91% COD and 80% TOC removals are obtained at the reaction temperature of 140 {sup o}C, the air total pressure of ...

Energy Technology Data Exchange (ETDEWEB)

Since the first commercial operation in 1983, Wolsung unit 1 has experienced several aging phenomena, especially in pressure tubes and feeder pipes. In case of pressure tubes, in-service inspections (ISI) revealed that a major portion of inspected tubes was in contact with calandria tubes. The likelihood of blister formation was a safety concern because it is a potential threat to pressure tube integrity. Wolsung unit 1 was, therefore, subjected to the SLAR (Spacer Location And Reposition) work to separate the contacted pressure tubes from calandria tubes. In this paper, experience of in-service inspection to the pressure tubes will be discussed including the irradiation creep elongation and CIGAR (Channel Inspection and Gauging Apparatus for Reactors) examination of pressure tubes. On the other hand, the problem of wall thinning in the ...

Energy Technology Data Exchange (ETDEWEB)

The performance of bunch-feedback systems based on two-tap FIR filters is discussed, particularly for transverse cases. Since they are the simplest type of FIR digital filters, the two-tap FIR filters should be suited to systems that require a very fast processing speed. However, in these feedback systems, a bunch must turn along a ring several times from the position-measurement to feedback kicking. Within these turns, irregular phase-jumps in the betatron motion can occur, due to several causes, such as the wake forces and beam-beam force. The discussion given in this paper is focussed on the effect of beam-beam collisions on these feedback systems. Based on linear theory and simulation studies, it has been found that these feedback systems can work with good performance, even under the influence of a strong beam-beam force. (author)

International Nuclear Information System (INIS)

Using the Laser-Fluorescence technique, an investigation was made, adopting the standard mix method, on trace uranium concentrations in hot spring water and geothermal water from Lishan region, and in tap water from some major cities in Shanxi province. Totally 40 samples from 27 sites were investigated. Measurement showed that the tap water contains around 10"-"6 g/L of uranium, whose concentrations in both hot spring water and geothermal water are 10"-"5 g/L. Most of samples are at normal radioactive background level, some higher contents were determined in a few samples

Energy Technology Data Exchange (ETDEWEB)

A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section ...

International Nuclear Information System (INIS)

This paper presents actual time/motion data for an oversize truck spent-fuel shipment from its origin, Surry, Virginia to its destination, Idaho National Engineering Laboratory (INEL). These data include the receipt of the empty cask at the reactor, wet-loading the cask, over-the-road or in-transit data, and receipt and dry unloading of the shipping cask at the receiving facility. Occupational doses were recorded at the Surry Power Plant as well as at INEL, and public doses were calculated for the in-transit dose analysis. This shipment was one of a series performed in support of a demonstration and evaluation of dry storage at INEL. The oversized shipment consisted of a TN-8L shipping cask loaded with three 10-yr-old pressurized water reactor assemblies. The total distance traveled was #approx#2800 miles, requiring 62 h including stops. The time required to receive and inspect the empty shipping cask and wet-load and ...

Science.gov (United States)

The Advanced Coal Liquefaction Research and Development Facility in Wilsonville, Alabama, is funded by the US Department of Energy (DOE), the Electric Power Research Institute (EPRI), and Amoco Corporation. On behalf of these organizations, Southern Company Services manages and Southern Clean Fuels Division of Southern Electric International operates the Wilsonville facility. Oak Ridge National Laboratory (ORNL) receives funding from DOE to provide materials technical support to the Wilsonville operators. For the period July 1987 through November 1990 the plant was operated with two reactors a thermal reactor and a catalytic reactor in a close-coupled integrated two-stage liquefaction mode. Coal processed was obtained from several seams including Ohio No. 6, Illinois No. 6, and Pittsburgh No. 8, as well as Texas lignite and several subbituminous coals. Corrosion samples which were removed for examination at the end of this ...

Energy Technology Data Exchange (ETDEWEB)

The Feedwater System supplies feedwater to the steam generator at the required pressure, temperature and flow rate during the plant start-up, normal power operation, shutdown. When the Feedwater System is inoperable or unavailable, the Auxiliary Feedwater System supplies emergency feedwater to the steam generator. If main steam line break occurs, the increase of feedwater flow rate of the faulted steam generator due to decrease of the pressure in the faulted steam generator results in adverse effects in aspect of overcooling the Reactor Coolant System and increased containment pressure/temperature. To optimize the containment mass/energy analysis, this paper evaluates the maximum feedwater and auxiliary feedwater flow rate delivered to the faulted steam generator at each stage of pressure decrease in the faulted steam generator after a main steam line break accident. Calculated ...

UK PubMed Central (United Kingdom)

Total antioxidant performance (TAP) measures antioxidant capacities in both hydrophilic and lipophilic compartments of serum and interactions known to exist between them. Our objective was to assess...Full Text Available

UK PubMed Central (United Kingdom)

A retrospective study is presented of 83 athletes with tendo Achillis pain (TAP) treated conservatively over a 12-year period from 1976 to 1988. Local steroid injections did not contribute to an earlier...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

Rats were exposed to lead (Pb) before or after weaning. For preweaning exposures, nursing dams received 0.2% Pb acetate(1090 ppm Pb) in drinking water. Offspring treated after weaning consumed the same concentration. Tapwater served as the control fluid. Pre-post treatments were distributed among four experimental groups of 10 rats each: tap water-tap water, Pb-tap water, tap water-Pb, Pb-Pb. Operant behavior training began on Postnatal Day 58. Animals performed on a multiple reinforcement schedule of food presentation consisting of fixed-interval, fixed-ratio, and time out components.Statistical analyses of experimental sessions 4 to 11 (Postnatal Days 72 to 79) revealed that postweaning exposure significantly lengthened the median interval (interresponse time) between consecutive responses on both the interval and ratio schedules. Preweaning exposure alone tended to produce a shortening of the median ...

Energy Technology Data Exchange (ETDEWEB)

A contamination event in a water distribution system can result in adverse health impacts to individuals consuming contaminated water from the system. Assessing impacts to such consumers requires accounting for the timing of exposures of individuals to tap-water contaminants that have time-varying concentrations. Here we present a probabilistic model for the timing of ingestion of tap water that we developed for use in the U.S. Environmental Protection Agency's Threat Ensemble Vulnerability Assessment and Sensor Placement Tool, which is designed to perform consequence assessments for contamination events in water distribution systems. We also present a statistical analysis of the timing of ingestion activity using data collected by the American Time Use Survey. The results of the analysis provide the basis for our model, which accounts for individual variability in ingestion timing and provides a series of potential ingestion times for ...

Energy Technology Data Exchange (ETDEWEB)

The corrosion resistance of steel-reinforced mortar specimens made from alinite cement was investigated using ordinary Portland cement (OPC) specimens as reference. The specimens were prepared and exposed in three different environments: continuous exposure in tap water, interrupted exposure in tap water, and interrupted exposure in 3.5% NaCl solution. The steel weight loss and the half cell potential were measured vs. exposure time, up to the age of 12 months. Pore solution extraction and analysis and porosity determination were also performed. In continuous exposure in tap water, alinite cement provided adequate protection against corrosion. In interrupted exposure in tap water, a higher corrosion was observed for alinite cement compared to OPC. In the case of interrupted exposure in 3.5% NaCl solution, the simultaneous action of free chlorides and oxygen resulted in the depassivation of steel ...

International Nuclear Information System (INIS)

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both ...

International Nuclear Information System (INIS)

The paper presents results of measurements of the natural radioactivity of tap water where samples were taken in front of, and behind the ceramic filter commonly used in houses for the purification of tap water. Altogether, 289 samples were taken, processed and measured during 1985-1986 in Kuwait. Results reveal the fact that ceramic filters reduce substantially the natural radioactivity in water (the 'gross' alpha activity reduced by the factor 2.18 #+-# 18%; the 'gross' beta by 1.53 #+-# 1.6%. (author).

Energy Technology Data Exchange (ETDEWEB)

Since 1976, the Nuclear Engineering Laboratory of the Technical Research Centre of Finland and Lappeenranta University of Technology have cooperated in the field of nuclear reactor thermal-hydraulics. During these years, a series of experimental facilities (REWET-I, -II, -III, VEERA) simulating pressurized water reactors (PWRs) have been built. The newest facility, PACTEL (Parallel Channel Test Loop), is an experimental out-of-pile facility designed to simulate the major components and system behaviour of a commercial PWR during postulated small and medium size break loss-of-coolant accidents (LOCAs), natural circulation and operational transients. A PACTEL natural circulation experiment has been carried out as an OECD/NEA international standard problem ISP 33. (2 refs., 3 figs., 2 tabs.).

Energy Technology Data Exchange (ETDEWEB)

The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events.

British Library Electronic Table of Contents (United Kingdom)

Strut lattice structures of reaction-bonded silicon infiltrated silicon carbide ceramics (RB-SiSiC) for air-fuel mixture formation and for nonstationary lean-burn under pressure applications were fabricated. The lattice design with a high porosity >80% was shaped by indirect three-dimensional printing. It was shown that pre-ignition processes in the porous reactor are much faster than in a free combustion, especially at lower temperatures. Interaction of high velocity diesel jets with cylindrical strut ligaments of the SiSiC lattice structure offers a new possibility for quick and efficient fuel distribution (multi-jet splitting) in space.

International Nuclear Information System (INIS)

The use of heat transfer augmentation devices has been proposed as a means for reducing the cost of steam generators for advanced LMR designs by approximately 25%. Experimental results are presented for two types of enhancers in the form of a twisted-tape and a core-tube insert in the steam tube of a 13.1-m-long, sodium-heated single-tube test section. Test parameters were prototypical of liquid metal reactor (LMR) steam generator design for a once-through high-pressure steam cycle. Heat transfer results are compared with the no-insert case on an overall and local basis. Although both augmentations increased overall performance, the mechanisms were different.

Energy Technology Data Exchange (ETDEWEB)

The Power Systems Development Facility (PSDF) is a joint Department of Energy/Industry sponsored engineering-scale facility for testing advanced coal-based power generation technologies. High temperature, high pressure gas cleaning is critical to many of these advanced technologies. Barrier filter elements that can operate continuously for nearly 9000 hours are required for a successful gas cleaning system for use in commercial power generation. Since late 1999, the Kellogg Brown & Root Transport reactor at the PSDF has been operated in gasification mode. This paper describes the test results for filter elements operating in the Siemens-Westinghouse particle collection device (PCD) with the Transport reactor in gasification mode. Operating conditions in the PCD have varied during gasification operation as described elsewhere in these proceedings (Martin et al, 2002).

Energy Technology Data Exchange (ETDEWEB)

Natural circulation experiments carried out in the REWET-III facility in 1985 have been used for RELAP5/MOD2 assessment. The REWET-III facility is a scaled-down model of VVER-440 type reactors. The facility consists of a pressure vessel in which the downcomer is simulated with an external pipe assembly, hot and cold legs with loop seals and a horizontal steam generator. The volume scaling factor compared to the reference reactor is 1:2333. The present paper summarizes the experiences gained in the RELAP5/MOD2 calculations of selected REWET-III single- and two-phase natural circulation experiments. The code's ability to represent the main phenomena of experiments in both cases was satisfactory.

International Nuclear Information System (INIS)

Natural circulation experiments carried out in the REWET-III facility in 1985 have been used for RELAP5/MOD2 assessment. The REWET-III facility is a scaled-down model of VVER-440 type reactors. The facility consists of a pressure vessel in which the downcomer is simulated with an external pipe assembly, hot and cold legs with loop seals and a horizontal steam generator. The volume scaling factor compared to the reference reactor is 1:2333. The present paper summarizes the experiences gained in the RELAP5/MOD2 calculations of selected REWET-III single- and two-phase natural circulation experiments. The code's ability to represent the main phenomena of experiments in both cases was satisfactory.

Energy Technology Data Exchange (ETDEWEB)

A descriptive model and design procedure for the DC electromagnetic flow coupler is developed based on a quasi-one-dimensional analysis previously developed for the DC electromagnetic pump. It is shown that for a particular flow coupler geometry, the total efficiency and the pressure gradients through the pump and generator depend on two parameters - the Hartmann number and the ratio of the pump flow rate to generator flow rate. Thus, for a fixed Hartmann number the efficiency depends only on the flow ratio. However, for a fixed pressure rise through the pump it is shown that the efficiency depends only on the Hartmann number. Nomographs showing the operating characteristics and critical design points are presented. Example calculations for a full-size unit, suitable for use in a liquid-metal cooled fast breeder reactor, are also discussed using the design nomographs.

International Nuclear Information System (INIS)

The helium releases were investigated in FBR fuel pins irradiated to high burn-up. The released amounts increased with the increase of burn-up, but no burn-up dependence of their release fraction to total amounts generated was seen. The contents of "2"4"1Am in fuel pellets affected the released amounts. The effect of released helium on the internal pressure in fuel pins was smaller in FBR fuel pins than in LWR fuel pins, but released helium in fuel pins containing increased amounts of "2"4"1Am, which are anticipated in future FBRs, is expected to affect the internal pressure.

Science.gov (United States)

Plastic composites have become a large class of construction material for exterior applications. One of the main disadvantages of wood plastic composites resides in the weak adhesion between the polar and hydrophilic surface of wood and the non-polar and hydrophobic polyolefin matrix, hindering the dispersion of the flour in the polymer matrix. To improve interfacial compatibility wood flour can be pretreated with environmentally friendly methods such as cold-atmospheric pressure plasma. The objective of this work is therefore to evaluate the potential of plasma polymerization of acetylene on wood flour to improve the compatibility with polyolefins. This presentation will describe the reactor design used to modify wood flour using acetylene plasma polymerization. The optimum conditions for plasma polymerization on wood particles will also be presented. Finally preliminary results on the wood flour surface properties and use in wood plastic ...

Energy Technology Data Exchange (ETDEWEB)

In the present work, the process of carbon dioxide absorption is analyzed at high partial pressures, in aqueous solutions of 1-amino-2-propanol (monoisopropanolamine (MIPA)), in relation to the thermal effects involved. All experiments were made in a stirred-tank reactor with a plane unbroken gas-liquid interface. The variables considered were the MIPA concentration within the range 0.1--2.0 M and the temperature within the interval 288--308 K. From the results, the authors deduce that the absorption process takes place in the nonisothermal instantaneous regime and propose an equation which not only relates the experimental results of flow density with the initial concentration of amine but at the same time enables the evaluation of the rise in temperature in the gas-liquid interface.

Energy Technology Data Exchange (ETDEWEB)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on current design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex ...

International Nuclear Information System (INIS)

The cold-worked Zr-2.5%Nb alloy is used as material for the pressure tubes in CANDU nuclear reactors. During the service life in reactor, diffusion of hydrogen and/or deuterium in the pressure tubes wall occur. Below a certain temperature, a stable hydride of zirconium is formed, as a brittle phase which can lead to catastrophic failures. For this reason, it is very important to be able to investigate the hydrogen effect on the micro structural properties of zirconium alloys. In the present paper a non-destructive testing technique is used, known as ultrasonic spectral analysis. When an ultrasonic signal traverses a medium, the frequency components associated with the input signal are altered. By frequency analysing the reflected signals, it is possible to study and compare the material properties. The two major parameters measured in ultrasonic spectroscopy are the attenuation and the velocity of the ...

International Nuclear Information System (INIS)

... computer calculations fftf reactor nonlinear problems reactor accidents reactor

Science.gov (United States)

Solid solution-strengthened nickel base alloys have been used for nuclear stream generator tubing in pressurized water reactors since the beginnings of commercial nuclear power. The purpose of this paper is to recap and update the authors understanding of the relationship between processing, resulting structure, and properties for Alloy 600 and to discuss the requirements for optimized performance in both primary and secondary environments. Potential replacement materials and their performance will be discussed. Also discussed is the role and importance of system chemistry, bulk and local, and control and its relationship to performance. A discussion of potential mechanisms of environmentally assisted failure is also discussed.

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code to design a pressurized water reactor and to obtain the licenses including entire proprietary rights is under development in domestic R and D project. The tasks of KAERI is to develop the constitutive relations including models for defining flow regimes and flow regime related models for inter-phase friction, wall frictions, wall heat transfer, and interphase heat and mass transfer in the two-phase three-field equations. In this paper, the process will be presented for choosing the best flow regime maps which occur in gas-liquid two-phase flow in horizontal and vertical tubes.

International Nuclear Information System (INIS)

An ion chromatograph has been modified to automatically sample ten liquid water streams from the secondary side of three pressurized water reactors, Calvert Cliffs, Unit One, Rancho Seco and McGuire, Unit 1. Sampling and measurement is semicontinuous with a cycle time of approximately five hours for 10 locations with sensitivities in the range of 0.1 to 0.5 ppb. The efficiency of the condensate polishing system and subsequent transport of sodium, chloride, and sulfate around the system can be readily followed. Sulfate has been shown to have unusual volatility into the steam phase from the steam generator as well as a tendency to pass through the condensate polisher.

Energy Technology Data Exchange (ETDEWEB)

Mitsubishi is now developing a new Pressurized water reactor (PWR), the Mitsubishi simplified PWR (MS-PWR), which has the innovative features of hybrid safety systems (an optimum combination of passive and active systems) and cooling by horizontal steam generators. In order to confirm the feasibility of the Mitsubishi hybrid safety system, various kinds of safety analyses are performed for loss-of-coolant accident events. In parallel to these safety analysis efforts, the following thermal-hydraulic tests are to be performed: (1) thermal-hydraulic test of a horizontal steam generator; (2) integrated thermal-hydraulic test using a simulation loop for the innovative MS-PWR (SLIM).

International Nuclear Information System (INIS)

The present process allows to eliminate the corrosion products formed on the tube plates and in the interstices of plate-tube crosspieces of a PWR steam generator in order to avoid a corrosion phenomenon which may cause denting by presence of oxides. The process consists in applying on these oxides at about 50-100 degrees, an aqueous solution containing 6-8% of gluconic acid, 3-5% of citric acid, about 0.5% of a corrosion inhibitor and ammonia until a pH of 3-9.5 is obtained.

Science.gov (United States)

Eight years ago, corrosion and tube denting seriously threatened the reliability and design life of steam generators, especially for closed loop arrangements in pressurized water reactors (PWRs). Concentrated research by the Steam Generator Owners Group (SGOG) diagnosed the causes and produced effective solutions, notably guidelines for water chemistry control in the secondary loop. The guidelines recommend specific levels of water impurities and remedial actions to prevent cooling-water leaks in the condenser, prevent air leaks, limit corrosion product buildup, and remove some impurities while neutralizing others. Continued research in SGOB-II is investigating intergranular corrosion and stress corrosion cracking. 3 figures.

International Nuclear Information System (INIS)

In a supercritical water-cooled reactor, property of water changes significantly around the critical point. It is expected that irradiation and change of water property will affect the chemistry and material corrosion. Deep understanding of interactions between supercritical water and materials under irradiation is important. However, comprehensive data on radiolysis, kinetics, corrosion and thermodynamics have not been obtained due to the severe experimental condition. To get such data by experiments and computer simulations, a national program funded by Ministry of Education, Culture, Sports, Science and Technology (MEXT) has been started since December 2002. (author)

International Nuclear Information System (INIS)

A safety analysis code, named as SPACE, for a pressurized water reactor is under development to obtain a licensing to be used for the PWR design and to hold entire proprietary rights. The task of KAERI is to develop the physical models and correlations which are required to solve the field equations. It can be divided into four parts; i) flow regime determination, ii) wall heat transfer, iii) wall and interfacial friction, iv) interfacial heat and mass transfer. This paper will describe the process to develop the models for the two-phase flow regime maps in the horizontal and vertical pipes.

International Nuclear Information System (INIS)

Plans, engineering parameters, and some test results for several FFTF core and primary sodium circuit instrument systems are presented. The systems discussed include temperature, flow, pressure, leak detectors, level sensors, fuel failure monitoring, sodium impurity analysis and cover gas monitors. Since many of these instruments are similar to those used in other fast reactors around the world, only a brief description is presented for these systems. Results of recent demonstration tests of the FFTF Under-Sodium Viewing and Ranging system are also presented. (U.K.).

International Nuclear Information System (INIS)

This paper reports several design options for improved passive circulation flow investigated for use in small, modular liquid metal cooled reactors (LMRs). The purpose is to enhance the transition to natural convection cooling following loss of forced circulation flow, reducing thermal transients experienced by the fuel and possibly eliminating the need for emergency pony-motor flow. Design details to minimize pressure drops may also enhance maximum equilibrium power levels possible under natural circulation only.

International Nuclear Information System (INIS)

Unit one of Daya Bay, China's first nuclear power plant was officially opened in February 1994. The nuclear island has been built by Framatome and is an improved version of the Gravelines 5 and 6, 900MWe Pressurized Water Reactors. Extra seismic protection has been included because of greater earthquake risk. The heat exchanger capacity has also been increased as the sea cooling water can be at 30"oC. The technical specifications and details of the fuel loading are given. The technical assistance, management and training of Chinese personnel are discussed. Two further units may be built if Daya Bay 1 is successful. (UK).

International Nuclear Information System (INIS)

Based on the occurrence of a number of plant incidents during low power and shutdown operating conditions, the Nuclear Regulatory Commission (NRC) has initiated several programs to better quantify risk during these periods. One specific issue of interest is the loss of residual heat removal (RHR) under reduced coolant inventory conditions. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-3 experimental facility at Siemens-KWU to supply applicable data. Recently, an effort has been undertaken at the Idaho National Engineering Laboratory (INEL) to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following loss of vital AC power and consequent loss of the RHR system during reduced inventory operation. The thermal-hydraulic response of a nuclear steam supply system (NSSS) with a closed reactor coolant system (RCS) to ...

Energy Technology Data Exchange (ETDEWEB)

On March 28, 1979, the Pennsylvania Three Mile Island nuclear power plant Unit 2 (TMI-2) suffered a partial fuel-melt accident. During this accident, over 20,000 lb of molten fuel flowed through holes melted through the baffle plates and through the lower-core support assembly (LCSA). The molten fuel subsequently resolidified in the bottom of the reactor vessel. The lower-core support assembly of the TMI-2 reactor was not structurally damaged during the accident. In order to permit defueling of that region of the core, the LCSA was cut to permit access. A five-axis teleoperator was developed to deliver plasma arc cutting, rotary grinding and abrasive waterjet cutting of end effectors to the LCSA. Complex geometry sectioning was completed in a mock-up facility at chemistry and pressure conditions simulating those of the vessel, prior to actual in-vessel operations. In-vessel activities began in early May 1988 and were ...

Energy Technology Data Exchange (ETDEWEB)

The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power Plants.`` The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell is used for the remote disassembly of irradiated fuel and material experiments. The absorber open-test assembly (AOTA) is a 12-m (40-ft)-long instrumented absorber (control-rod-material) test assembly. Its primary purpose is to characterize the FFTF control-rod-material reaction rate during reactor operation. Instrumentation allowed temperature and pressure measurements at various locations in several absorber pins during reactor operation. After residing several months in the reactor, the assembly was transferred to the IEM cell by the closed-loop ex-vessel machine (CLEM) for separation of the irradiated portion of the experiment from the instrument stalk. After separation, the 3.6-m (12-ft)-long assembly was processed through the sodium removal system and shipped off-site for examination. This success allowed the timely ...

International Nuclear Information System (INIS)

Hydrogen is a potential alternative energy source and produced commercially by methane (natural gas) or LPG steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, since this process emits large amounts of CO_2, replacement of the combustion heat source with a nuclear heat source for 773-1173 K processes has been proposed in order to eliminate these CO_2 emissions. This paper proposes a novel method of low-temperature nuclear hydrogen production by reforming dimethyl ether (DME) with steam produced by a low-temperature nuclear reactor at about 573 K. The authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573 K. By setting this low-temperature hydrogen production process at about 573K upstream from a turbine, it was found theoretically that the total energy utilization efficiency is about 50% and very high. By setting ...

Energy Technology Data Exchange (ETDEWEB)

On February 1, 1984, the US Nuclear Regulatory Commission issued Generic Letter 84-04 on the subject of postulated pipe breaks in pressurized water reactor (PWR) primary coolant loops, opening the way for pipe-whip restraint exemptions. The letter substitutes the leak-before-break (LBB) criteria for the double-ended guillotine break regarding PWR primary reactor coolant system (RCS) piping and asymmetric blowdown loads. The LBB criterion refers to the fact that a piping flaw will leak before it breaks. The current requirement to provide pipe-whip restraints is applied within the plant to all high-energy piping with a potential for damaging structures, systems, and components essential to safe reactor shutdown. This includes primary RCS piping 30 in. and larger as well as smaller piping systems. A study was performed to evaluate the applicability of the LBB criteria proposed in NUREG-1061 to the latter ...

Energy Technology Data Exchange (ETDEWEB)

In the hypothetical event of a severe accident in a Light Water Reactor, scenarios in which the reactor pressure vessel (RPV) fails and the core melt mixture (called corium) relocates into the reactor cavity, cannot be excluded. The viscosity (in fact, corium rheological behaviour) plays a major role in many phenomena such as core melt down, discharge from reactor pressure vessel, interaction with structural materials (concrete,...) and spreading in a core-catcher. For these reasons, it is important to be able to predict the rheological behaviour of corium melts of different compositions (essentially based on UO{sub 2}, ZrO{sub 2}, Fe{sub x}O{sub y} and Fe for in-vessel scenarios, plus SiO{sub 2} and CaO for ex-vessel scenarios) at temperatures above solidus temperature. In the case of corium-concrete mixtures, the increase of viscosity depends not only on the ...

International Nuclear Information System (INIS)

Previous flow regime studies have been for horizontal, vertical, and inclined pipe flow. As such, only a few studies have been performed on bundle geometries. The present paper examines the flow regimes for a vertical hexagonal flow channel with and without a finned fuel bundle. This type of a 36 finned rod hexagonal fuel bundle in parallel hexagonal flow channels is used in a MAPLE (Multi- purpose Applied Physics Lattice Experimental) type nuclear reactor. An experiment apparatus was designed consisting of the flow channel, inlet plenum and an air-water separator. The inlet plenum is used to provide a uniform mixture of air and water before entering the hexagonal flow channel. A turbine flow meter is used to determine the water flow rate. The turbine flow meter is calibrated for a low flow range and limits the measurable flow to 50 l/min. Flow pattern observation is determined by a SONY video camera, Real-Time Neutron Radiography, pressure ...

Energy Technology Data Exchange (ETDEWEB)

This semiannual report for the period October 1983-March 1984 summarizes activities in Sandia National Laboratories' continuing program of coal liquefaction research. The primary goals are to: explore novel catalytic concepts and materials for conversion of coal to liquid fuels; determine the effects of process variables on catalyst deactivation; determine the effects of coal structure and solvent properties on low temperature dissolution; study the kinetics and catalysis of hydrogen transfer reactions; develop an understanding of slurry gelling phenomena; and provide a technical assessment of coal liquefaction processes. During this period, work was performed on: the use of pyrene as a chemical probe of catalyst activity; analysis of catalysts from Wilsonville run 242 using ESCA; atmospheric pressure model compound activity testing of regenerated catalysts from Wilsonville run 242; base displacement experiments with a coal-indole complex; preliminary ...

International Nuclear Information System (INIS)

CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO_2), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel ...

International Nuclear Information System (INIS)

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor ...

Science.gov (United States)

This thesis study covers both global performance and local phenomena analyses focusing on natural circulation reactor design safety. Four important topics are included: the global SBWR design safety assessment, important local phenomena investigation, steady and transient natural circulation process study, and two-phase instability analysis. The conceptual design of the SBWR-200 is introduced in this thesis and the global performance of a natural circulation reactor is then assessed using PUMA integral test data and RELAP5 simulations. A safety assessment methodology is developed to evaluate the PUMA integral test data extrapolation and code scalability. The RELAP5 code simulation capability in low-pressure low-flow conditions is also validated. The study shows that the code is capable of predicting the global accident scenario in natural circulation reactors with reasonable accuracy, while failing to ...

Energy Technology Data Exchange (ETDEWEB)

The Control Rod Drive(CRD) nozzles for PWR nuclear power plants(NPP) house the control rod drives. The number of nozzle penetrations range from the mid-30's to over 100 in each reactor head. The integrity of CRD nozzles is very important, because the primary pressure boundary is established with the J-groove weld joining the nozzle to the head clad surface. The Alloy 600 PWSC CRD nozzle leaks discovered in the fall of 2000 and spring of 2001 in several US plants. Therefore the NRC has recommended a more proactive effort by US utilities to inspect similarly susceptible nozzles in all US plants. The primary safety concern is circumferential cracks that can permit the nozzles to separate from the head at high velocity and produce a large-break leak in the reactor vessel. A secondary concern is head leakage from any through-wall cracks in the nozzle or J-groove weld area. Numerous inspection and repair tools have been ...

Energy Technology Data Exchange (ETDEWEB)

Experiments were performed to measure the transfer of trichloroethylene (TCE), a volatile organic compound (VOC), from tap water in showers to indoor air. In these experiments, the loss of TCE from tap water in the shower is based on the difference between influent and effluent concentrations.We have developed and previously published a three-compartment model, which we use to simulate the 24-h concentration history of VOCs in the shower, bathroom, and remaining household volumes resulting from the use of contaminated tap water. An important input to this model is the transfer efficiency of the VOC from water to air. The experiments reveal that the transfer efficiency of TCE from shower water to air has an arithmetic mean value of 51 percent and an arithmetic standard deviation of 9 percent. Analysis of the results shows that there is no statistically significant difference between the transfer efficiency measured with hot ...

Energy Technology Data Exchange (ETDEWEB)

This paper elucidates morphological characteristics, reproduction forms and ecological characteristics of Plectus sp. and Rhabditis sp. which are most frequently detected nematodes separated from tap water. The paper also discussed experimentally establishment of cultivating conditions for these nematodes, and behavior and reproduction characteristics of nematodes under widely ranging conditions of temperatures, bacteria concentrations and pH. Plectus sp. grows in hermaphroditic individuals, while Rhabditis sp. grows only in female. An LE culture medium can cultivate nematodes that appear in tap water, and is suitable for subculture of nematodes. Conditions of temperatures from 20{degree}C to 25{degree}C, pH and inoculated concentration of bacteria at 10{sup 8} cell {center_dot} ml{sup -1} are suitable for reproduction of Plectus sp. and Rhabditis sp. Plectus sp. and Rhabditis sp. prefer mild temperatures for reproduction, cannot reproduce ...

International Nuclear Information System (INIS)

The paper introduces the disposal procedures of the old reactor pressure vessel head. Reactor pressure vessel (RPV) heads were replaced successfully in Daya Bay Nuclear Power Plant. The on-site treatment and disposal of replaced heads was thereafter implemented. This is the first time in China for large size radioactive article disposal. To guarantee the safety of old head treatment, issues of radiation protection have to be taken into account. Dose rate distribution of the old head surface and radioactivity were initially calculated according to the operation history and the neutron flux rate of the reactor. Shielding calculation and package design was thereafter worked out based on the data obtained from the old head, e.g. neutron flux, sorts of nuclides, radioactivity and dose rates. Shielding package was afterwards manufactured in special factories and transported to Daya Bay ...

Energy Technology Data Exchange (ETDEWEB)

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25 m x 0.05 m) and 2.59 m, respectively, whereas the inclination angle of the riser is 50 deg. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels measured in the separators. The counter-current flow limitation is defined as the maximum air mass flow rate at which the discharged ...

International Nuclear Information System (INIS)

Inputs on radiological conditions forms the basis of implementation of effective exposure control to plant personnel in nuclear power station. Radiation monitoring system provides this input to the plant operator as well as to health physics group. Several design modifications have been incorporated in the Radiation Monitoring System at Tarapur Atomic Power Station (TAPS 3 and 4) over the similar systems at Kakrapar Atomic Power Station (KAPS) and Kaiga Generating Station (KGS). The radiological monitoring systems installed at TAPS unit 3 and 4 includes on line Radiation Data Acquisition System (RADAS), Emergency sampling system, effluent monitoring system and environmental monitoring system. The design changes and the versatile use of these systems are presented in this paper. (author)

Energy Technology Data Exchange (ETDEWEB)

The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the ...

Energy Technology Data Exchange (ETDEWEB)

Liquid metal cooling for the first wall and blanket of a magnetic confinement fusion reactor has various advantages. However, it has the disadvantages of large magnetohydrodynamic pressure drops and heat transfer deterioration under a strong magnetic field. Thus, the present authors have proposed cooling with a helium-lithium annular mist flow as well as the cooling with a liquid metal boiling flow, and as fundamental studies, investigated the effect of a magnetic field on the flow characteristics and heat transfer of liquid metal two-phase systems since the 1970s. In the present paper we summarize the important findings obtained from our experimental studies for (i) an air-mercury stratified flow in a horizontal rectangular channel, (ii) a helium-lithium annular mist flow in a horizontal rectangular channel, (iii) the mercury pool boiling on a horizontal surface, and (iv) air-mercury upward flows in a vertical circular tube. Based on the ...

International Nuclear Information System (INIS)

Commissioning of a Pressurized heavy water reactor (PHWR) plant of NPCIL involves three phases viz phase-A which consist of pre-criticality activities such as hydro test, air hold test, no load test of motors etc., phase-B consist of criticality and post criticality physics experiments. The phase-C, which is considered the major phase, consist of initial power raise to about 10 % , TG rolling, synchronization, going to significant power in steps and performance tests such as load rejection tests from various power levels. In order to have smooth commissioning for the Phase-C, an integrated team consisting of engineers from various design and analysis groups of NPCIL headquarters was formed to participate along with site O and M engineers, closely observe and coordinate phase-C commissioning activities. During this commissioning some major events and observations took place. An attempt is made to bring out the salient observations of Tarapur-4 ...

Science.gov (United States)

Because hydrogenation of coal to liquid products (oils) is accompanied by distributions of complex by-product mixtures (IOM, preasphaltenes, asphaltenes and gases) which change as a function of reaction variables (time, temperature and pressure) and reactor configuration, the determination of selectivity relationships for coal liquefaction catalysts has been a difficult and time-consuming task involving numerous experiments to adequately describe catalyst performance over a range of conditions. This paper describes a method for analyzing the experimental results of coal liquefaction reactions which may be applied to a number of aspects of coal liquefaction research and process control, including: rapid selectivity and performance screening for catalysts; correlation of laboratory results with process parameters; and optimization of product yield for plant process conditions. Catalyst selectivity and performance screening will be emphasized ...

International Nuclear Information System (INIS)

This paper presents part-load evaluation of a natural gas-fired chemical looping combustion (CLC) combined cycle with CO2 capture. The novel combined cycle employs an air-based gas turbine, a CO2-turbine and a steam turbine cycle. In this combined cycle, the CLC reactors replace combustion chamber of the gas turbine. The proposed combined cycle has a net plant efficiency of about 52.2% at full-load, including CO2 compression to 200 bar. The part-load evaluation shows that reducing the load down to 60% results in an efficiency drop of 2.6%-points. However, the plant shows better relative part-load efficiency compared to conventional combined cycles. The pressure in CLC-reduction and -oxidation reactors is balanced by airflow control, using a compressor equipped with variable guide vanes. A combination of control strategies is discussed for plant start-up and shutdown and for part-load when airflow reduction is not ...

Energy Technology Data Exchange (ETDEWEB)

SiC has been considered as a primary candidate material for a first wall component in future fusion reactor because it has been claimed that SiC has excellent high-temperature properties, good chemical stability and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, tritium trapping capacity on the surface of SiC was experimentally obtained at the temperature range of 25-800 deg. C in consideration of tritium trapping to the experimental system. The capacity, which was independent of the water vapor pressure in the gas phase and the temperature, was determined as about 10{sup 6} Bq/cm{sup 2}. The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface was quantified at the temperature of 25, 500 and 700 deg. C in consideration of the behavior of tritium trapping at change of experimental condition by the numerical curve fitting method applying the serial ...

Energy Technology Data Exchange (ETDEWEB)

The NUclear Power Engineering Corporation (NUPEC) has developed IMPACT-SAMPSON code to analyze integral behavior of light water nuclear power plants under severe accident conditions. IMPACT-SAMPSON's distinguishing features include interconnected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to severe accident events, and high-speed simulation on parallel processing computers. The integral plant behaviors of typical PWR and BWR under severe accident conditions have been analyzed with the IMPACT-SAMPSON code. The PWR plant analyzed was the three-loop, steel-dry containment type with 2,440 MWt. The AE accident scenario was supposed, that is, LOCA by 6-inch hot leg failure followed by accumulated water injection, but no ECCS and containment spray activation. The BWR plant analyzed was the 3,293 MWt BWR-5, Mark-II containment type. The TQUV accident scenario was supposed, that is, reactor scram ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford Site. The FFTF is the only reactor of this type designed and operated with the intent of meeting the licensing requirements of the Nuclear Regulatory Commission (NRC). Unique characteristics of the FFTF that present special challenges related to leak rate testing include thin wall containment vessel construction, cover gas systems that penetrate containment, and a low-pressure design basis accident. The successful completion in 1986 of the third FFTF Integrated Leak Rate Test (ILRT) five days ahead of schedule and 10% under budget was a major achievement for the Westinghouse Hanford Company. The success of this operational safety test was due in large part to a special local area network (LAN) of three IBM PC/XT computers that monitored the sensor data, calculated the containment vessel leak rate, and displayed test results. The ...

International Nuclear Information System (INIS)

One of inherent characteristics of CANDU reactor is positive void reactivity in contrast to other pressurized light water reactors. During the large break loss of coolant accident, power pulse will be occurred during short time of early phase of accident due to positive void reactivity. However the duration of this power pulse is short, energy due to power pulse would be accumulated in the cladding material and will affect the peak cladding temperature or number of failed fuel elements. Recently, Canadian Nuclear Safety Commission (CNSC) indicated that the amount of void reactivity might be larger than the assumed values in safety analysis and this indication was based on the experimental data from ZED-2 facility. Based on that, the estimation of uncertainties due to the void reactivity during LBLOCA is the most important issue for CANDU safety analysis. In this study, a framework of uncertainty evaluation methodology for ...

International Nuclear Information System (INIS)

The generation of fusion power from the Fast-Liner Reactor (FLR) concept envisages the implosion of a thin (3-mm) metallic cylinder (0.2-m radius by 0.2-m length) onto a preinjected plasma. This plasma would be heated to thermonuclear temperatures by adiabatic compression, pressure confinement would be provided by the liner inertia, and thermal insulation of the wall-confined plasma would be established by an embedded azimuthal magnetic field. A 2- to 3-mu s burn would follow the approx. 10"4 m/s radial implosion and would result in a thermonuclear yield equal to 10 to 15 times the energy initially invested into the liner kinetic energy. For implosions occurring once every 10 s a gross thermal power of 430 MWt would be generated. The results of a comprehensive systems study of both physics and technology (economics) optima are presented. Despite unresolved problems associated with both the physics and technology of the FLR, a conceptual power ...

International Nuclear Information System (INIS)

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for evaluation of debris in-vessel retention. The debris coolability analysis module is developed for the accurate prediction of the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with a temperature distribution and failure evaluation of the vessel wall. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis. The calculation for spreading is compared with a water spreading experiment on the floor and the calculation for coolability is compared with a n-octadecane melting experiment in a rectangular vessel. The comparisons show capabilities for predictions of spearhead transportation in the debris spreading ...

International Nuclear Information System (INIS)

Ferritic/martensitic steel, modified 9Cr-1Mo steels have been used most extensively in the power generation industry throughout the world due to having superior high temperature properties such as high strength, creep resistance, and good microstructure stability. These steels are also the primary candidate for the RPVs(Reactor Pressure Vessels) of High Temperature Gas-Cooled Reactors. Currently, many studies has been conducted in laboratory-scale for mod. 9Cr-1Mo steels. However, there is a lack of the study on forged thick- section for RPVs. The differences in characteristics including the through thickness microstructure and mechanical properties between internal and external locations may occur during cooling after austenitization, because the thickness of RPVs is over about 200mm. Therefore, in order to use ferritic/martensitic steel as RPVs, a detailed assessment of the through thickness properties is needed. The ...

International Nuclear Information System (INIS)

This paper gives an overview of a multiyear joint research program being conducted at the University of New Mexico (UNM) with support from Sandia National Laboratories and GA Technologies. This research focuses on heat removal and fluid dynamics in flow regimes characterized by low pressure and low Reynolds number. The program was motivated by a desire to characterize and analyze cooling in a broad class of TRIGA-type reactors under: (a) typical operating conditions, (b) anticipated, new operating regimes, and (c) postulated accident conditions. It has also provided experimental verification of analytical tools used in design analysis. The paper includes descriptions of the UNM thermal-hydraulics test facility and the experimental test sections. During the first two years experiments were conducted using single, electrically heated rod in water and air annuli. This configuration provides an observable and serviceable simulation of a fuel rod ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this research was to determine the extent of damage that occurs when two pipes experience an impact event due to one whipping against the other. The research was conducted through experimental and analytical approaches. The former required the development of a specialized impact machine that could accelerate a whipping pipe with sufficient energy to cause failure of a target pipe that was heated and pressurized to Pressurized Water Reactor (PWR) conditions. Damage was measured in terms of crushing, bending, and failure. The results of the tests permitted the correlation between pipes of a certain size and the damage they could cause when impacting with a certain amount of known energy. These results were used to evaluate the pipe whip criteria in the Standard Review Plan 3.6.2-4. It was established that the criteria conditions did not fully represent the results obtained experimentally. An analysis ...

Energy Technology Data Exchange (ETDEWEB)

Westinghouse is currently conducting the pre-application licensing of the International Reactor Innovative and Secure (IRIS). One of the key aspects of the IRIS design is its safety-by-designTM philosophy and within this framework the PRA is being used as an integral part of the design process. The most ambitious risk-related goal for IRIS is to reduce the Emergency Planning Zone (EPZ) to within the exclusion area by demonstrating that the off-site doses are consistent with the US Protective Action Guidelines (PAGs) for initiation of emergency response so that the required protective actions would be limited to the exclusion area. As a first step, a model has been developed to provide a first order approximation of the Large Early Release Frequency (LERF) as a surrogate predictor of the off-site doses. A key-aspect of the LERF model development is the characterization of the possible paths of release. Four main categories have been historically pointed out: (1) ...

International Nuclear Information System (INIS)

The static thermophysical properties of the molten salt system like LiF-NaF-BeF_2 influence the design and construction of the fuel salt and coolant in the Molten Salt Reactor for the new generation. In this paper, the equation of state of the ternary system 0.15LiF-0.58NaF-0.27BeF_2, over the temperature range from 873.15K to 1073.15K at one atmosphere pressure, is described by using modified Peng-Robinson equation. The density of the ternary system is evaluated by this equation directly, and compared with the experimental data. Base on the equation of state, the other static thermophysical properties such as the enthalpy, entropy and heat capacity at constant pressure are evaluated by the fugacity coefficient and residual function methods respectively. The density calculated by Peng-Robinson equation is in highly agreement with the experimental data, and the enthalpy, entropy and heat capacity evaluated by such two ...

Science.gov (United States)

The static thermophysical properties of the molten salt system like LiF-NaF-BeF2 influence the design and construction of the fuel salt and coolant in the Molten Salt Reactor for the new generation. In this paper, the equation of state of the ternary system 0.15LiF-0.58NaF-0.27BeF2, over the temperature range from 873.15K to 1073.15K at one atmosphere pressure, is described by using modified Peng-Robinson equation. The density of the ternary system is evaluated by this equation directly, and compared with the experimental data. Base on the equation of state, the other static thermophysical properties such as the enthalpy, entropy and heat capacity at constant pressure are evaluated by the fugacity coefficient and residual function methods respectively. The density calculated by Peng-Robinson equation is in highly agreement with the experimental data, and the enthalpy, entropy and heat capacity evaluated by such two ...

International Nuclear Information System (INIS)

On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract ...

J-STORE (Japan)

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International Nuclear Information System (INIS)

During the last years one focus of German PWR safety analysis was boron dilution events with the potential of reactivity transients. Coolant with a low boron concentration could be collected in localized areas of the reactor coolant system e.g. by separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux- condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. The TUeV NORD SysTec was charged by German supervisory authorities with the assessment of the safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses shall demonstrate that boron dilution events cannot lead to recriticality of the core. Hence the boron concentration at ...

International Nuclear Information System (INIS)

A series of Model Tests of Embedment Effect on Reactor Buildings has been carried out by the Nuclear Power Engineering Corporation (NUPEC), under the sponsorship of the Ministry of International Trade and Industry (MITI) of Japan. Seismic response of an embedded reactor building is greatly affected by the non-linearity of the backfill soil. However, quite few experimental data have been obtained so far. The objective of this study is to qualitatively evaluate the non-linear behavior of the backfill soil through shaking table tests. Its effects to the seismic response of a reactor building constructed at a soft rock site can be made clear through the tests. Non-linear effects of the backfill soil on the seismic response of the embedded reactor building model were evaluated experimentally. Based on the sinusoidal and seismic wave excitation tests, the following conclusions were obtained regarding the ...

Energy Technology Data Exchange (ETDEWEB)

My assignment on the SP-100 Vacuum Vessel Vacuum System Team was to perform a transient pump down analysis for the vacuum vessel that will house the SP-100 reactor during testing. Pump down time was calculated for air and helium. For all cases the proposed vacuum system will be able to pump down the vessel within the required time. The use of a larger rotary piston pump (DUO250) improves the pump down time by 35 minutes and therefore should be considered. The 6-inch duct for the roughing line is optimal, however, because all cases are well below the 24 hour time frame, the 4-inch duct is sufficient. The use of the single turbomolecular pump during pump down is sufficient. A pump down with helium in the vessel and a helium inleakage delays the time to achieve the base pressure marginally and is acceptable.

International Nuclear Information System (INIS)

The ROP margin in a CANDU reactor is decreasing over time due to the Primary Heat-Transport System (PHTS) aging effect. Adjustment of the ROP trip setpoint is required to maintain a high trip-probability and ROP trip effectiveness. Especially, for Wolsong-1, which is scheduled to change the old pressure tubes in 2009, the trend of ROPT after the retubing should be reevaluated. Before setting a ROPT, the main thermal characteristics including Critical Channel Power (CCP) should be calculated by the NUCIRC code. In this paper, the thermalhydraulic evaluation for Wolsung-1 was conducted with the updated Wolsung-1 PHTS data. Specifically, for the case of 0 EFPY (Effective Full Power Year) and 11 EFPY after the retubing, the distribution of the channel flow rate, channel exit quality, critical channel power, and critical power ratio (CPR) of the Wolsong-1 aged plant are calculated

Energy Technology Data Exchange (ETDEWEB)

Slow, unexpected displacements have been experienced in most pressurized water reactor (PWR) surge lines. Sometimes, these displacement lead to gap closure at the pipe whip restraints. These movements occur because of thermal stratification. This movement has the potential to increase stresses to valves, which may exceed the material yield stress. To understand this phenomenon, Framatome, Commissariat a l'Energie Atomique, and Electricite de France have undertaken large programs for the study of (1) thermal-hydraulic tests with a half-scale Plexiglas surge line, (2) thermal-hydraulic computations of permanent states and transients with a two-dimensional model, and (3) mechanical analysis of displacements and computation of fatigue damage due to stratification. This paper deals with the last subject. Avoiding stratification in piping by process modifications is difficult because of the high flow rate needed. Alternative solutions for ...

International Nuclear Information System (INIS)

The Surry low power and shutdown probabilistic risk analysis (PRA) is an ongoing project at Brookhaven National Laboratory (BNL) to identify and quantify potential accident scenarios that may occur in a pressurized water reactor (PWR) during low power and shutdown. It was initiated as a result of various incidents and accidents that have occurred within the United States and overseas. The project involves review and evaluation of PWR experience at shutdown, identification of accident scenarios, determination of methods to mitigate the accidents, and performance a level 1 PRA. An evaluation of accident progression, source terms and consequences has also been initiated. The results will be used to address issues related to shutdown conditions. The objective of this paper is to provide a progress report on the project, and to present the approach used as well as the preliminary results of the ongoing and completed tasks.

British Library Electronic Table of Contents (United Kingdom)

Porous flow occurs in a wide range of materials and applies to many commercially relevant applications such as oil recovery, chemical reactors and contaminant transport in soils. Typically, breakthrough and pressure curves of column floods are used in the laboratory characterization of these materials. These characterization methods lack the detail to easily and unambiguously resolve flow mechanisms with similar effects at the core scale that can dominate at the aquifer or oil field scale, as well as the effects of geometry that control the flow at interfaces as in a perforated well or the inlet of an improperly designed column. Non-invasive imaging techniques such as MRI have been shown to provide a far more detailed characterization of the properties of the solid matrix and flow, but usu...

Science.gov (United States)

In this paper we study theoretically a pulsed corona discharge plasma in air at atmospheric pressure. A self consistent one-dimensional model, using a flux-corrected transport numerical scheme, is presented to achieve it. The charged particle kinetics in this model are described by one-dimensional continuity equations coupled with Poisson's equations. The spatio-temporal local field, charge density variations and the velocity profile of an ionizing front are calculated to describe the dynamical behavior of corona discharge plasmas. The simulation results of a wire-in-cylinder corona discharge plasma explain the physical mechanisms of discharge processes. These results may also be apply to obtain the optimizing parameters for designing the plasma reactor.

Energy Technology Data Exchange (ETDEWEB)

It is shown that a gun similar to that in the SSPX could demonstrate the high fields required for Pulsed Spheromak reactors merely by prolonging the pulse. Important considerations are choosing the voltage to exceed ohmic losses; designing the gun to avoid wasteful short-circuiting of current within the gun; and the injection efficiency factor, f, determined by the ''sag'' in the profile of {lambda} = {mu}{sub o}j/B. Typically f = 0.75 in experiments, giving an overall efficiency > 50 % if short-circuiting is avoided. Theoretical transport models agree qualitatively with the need for a finite gradient in h to pump in helicity by current-driven tearing modes and suggest that pressure-driven resistive modes would not compete with current-driven modes during a buildup to ohmic ignition.

International Nuclear Information System (INIS)

The phenomenon of radiation blistering by helium ion bombardment has been the subject of extensive studies in recent years because of its technological importance in thermonuclear fusion devices and reactors. However, the mechanism of radiation blistering is still not well understood. There are two different models of blister formation: the gas-pressure model and the lateral stress model. The former model is, however, supported by many experimental observations, the prominent one is that of Evans and Eyre who observed blisters appearing on the front and rear surfaces of a thin wedge-shaped molybdenum foil irradiated by helium ions. Their experiment also indicates that the thickness of the irradiated specimen could be important in affecting the characteristics of blisters. With this in view, we have studied the development of blisters in thin foils of tantalum by 30 MeV #alpha#-particle bombardment. (orig.).

Energy Technology Data Exchange (ETDEWEB)

Explosives have rarely been used in decommissioning of nuclear reactors. Nevertheless, controlled blasting can be used advantageously during careful destruction of nuclear power plants for removal of concrete, pipe systems, and other components. Experiments performed within a former nuclear power plant demonstrate the feasibility of this method, employing explosive masses up to 15 kg per blast. The loadings of the components and the total plant structure were measured and compared with code predictions. The experiments show a response of the containment predominantly in frequency ranges above 100 Hz, thus keeping the building and components below German regulation limits for shock excitation. The blast wave pressures are reduced drastically within short distances in the building. Dust and debris can be contained with simple methods such as curtains. Use of this method seems to be applicable to actual dismantling projects.

Science.gov (United States)

The decomposition of three ?-lactam antibiotics (amoxicillin, oxacillin and ampicillin) in aqueous solution was investigated using a dielectric barrier discharge (DBD) in coaxial configuration. Solutions of concentration 100 mg/L were made to flow as a film over the surface of the inner electrode of the plasma reactor, so the discharge was generated at the gas-liquid interface. The electrical discharge was operated in pulsed regime, at room temperature and atmospheric pressure, in oxygen. Amoxicillin was degraded after 10 min plasma treatment, while the other two antibiotics required about 30 min for decomposition. The evolution of the degradation process was continuously followed using liquid chromatography-mass spectrometry (LC-MS), total organic carbon (TOC) and chemical oxygen demand (COD) analyses. PMID:21514950

International Nuclear Information System (INIS)

Zirconium alloys have anisotropic mechanical properties depending on their physical orientations and are widely used as nuclear materials such as cladding tube material. An operation condition of the nuclear reactor requires a high creep resistance, because it is subjected to long period operations, high temperature and high pressure. Generally, it takes a few days or months to do the creep experiment, so it is difficult to get a data in short period. However, there is a way to predict a creep property by using the stress-relaxation in the short term. These studies realized the stress-relaxation through a compressive test of HANA-6 (Zr-1.1Nb-0.05Cu) alloy that was developed by KAERI (Korea Atomic Energy Research Institute), and then predicted the creep property

British Library Electronic Table of Contents (United Kingdom)

A novel liquid phase alkylation of aniline with methanol over Zn/Cu salts modified kaolin assisted with a pair of porous carbon electrode in slurry-bed reactor under constant current intensity, room temperature and atmospheric pressure was reported. The Zn/Cu salts modified kaolin catalysts were synthesized and characterized by infrared spectrometer (IR), powder X-ray diffraction (XRD) and scanning electron microscopy (SEM), which showed that the transition metals were completely supported on kaolins structure and formed a pored one. The effect parameters, such as initial pH, electrolysis time, metal ratio with kaolin and salts composition in this electrochemical catalytic system, were studied. The procedure was inspected by ultraviolet-visible spectrum (UV-vis), and the product distributi...

International Nuclear Information System (INIS)

The Supercritical Water Cooled Reactor(SCWR) operates in a pressure around 25MPa and temperature of 293#approx#510 .deg. C. In order to study the heat transfer behaviors and good comparisons between the various fluids, a heat transfer test loop(SPHINX) using CO_2 has been constructed in KAERI as a part of international research program, I-NERI. At a supercritical pressure, the heat transfer coefficient is much larger than that estimated from the Dittus-Boelter correlation for a relatively large flow rate with moderate wall heat flux conditions. This phenomenon was explained by the rapid variations of the physical properties near the wall with the temperature. On the contrary, the heat transfer becomes worse when the bulk fluid enthalpy is below the pseudo-critical enthalpy under a low flow rate with large heat flux conditions. This phenomenon is called 'deteriorated heat transfer', and which is explained as the modification ...

Energy Technology Data Exchange (ETDEWEB)

Thermogalvanic corrosion of St3 steel in a temperature range from 25 to 50{degrees}C with a 3.84{degrees}C/cm gradient was studied in a neutral media flow. It is shown that the thermogalvanic effect on the pitting corrosion of water-cooled heat exchangers is decisive but dual. Depending on the composition of the cooling medium, thermogalvanic corrosion may either enhance (0.5 M NaCl) or inhibit (tap water) pitting corrosion.

International Nuclear Information System (INIS)

Optimum analytical conditions of the aluminium ion were established by flow injection analysis. Eriochrome Cyanine R(ECR) dye reacts with the aluminium ion at pH 6.0 form a complex that exhibits maximum absorption at 535 nm. Reaction condition including the mixing and the reaction coil length, the concentration and the pH of the buffer solution, temperature, and injection loop volume were optimized to introduce this reaction into flow injection analysis. The results were as follows. A mixing coil length of 0.5 m and a reaction coil length of 4.0 m, the pH 6.0 and 1M of acetate buffer solution, the ECR concentration of 0.56 mM, the reaction temperature of 40 .deg. C, the injection loop volume of 300 #mu#L were chosen as optimum conditions. Under these conditions the detection limit of the aluminium ion was less than 0.05 mg/L and the repeatability was better than 1%. A sampling frequency of 24 times for an hour was achieved. Interfering ions such as F"-, HPO_4"2"-, Fe"2"+, Fe"3"+, ...

International Nuclear Information System (INIS)

Despite successful reduction of the general lead exposure, this heavy metal is still a matter of public concern due to the fact that associations with intellectual impairment or delayed puberty are found to correlate with very low lead blood concentrations. Lead in tap water is still an important contribution to lead exposure which may cause health risk for infants. Therefore, lead pipes should be completely sanitated by exchange against pipes made from more healthy materials. (orig.)

International Nuclear Information System (INIS)

Full text of publication follows: The TUV NORD e.V. is an independent Technical Support Organisation (TSO) performing safety assessments in almost every field of technology. In nuclear safety the TUV can look back on more than 40 years of experience. In the last years in Germany PWR safety analyses were focussed on boron dilution events with the potential of reactivity transients. The possibility of coolant with a low boron concentration collected in localized areas of the reactor coolant system (RCS) can be caused by injection of coolant with less boron content from interfacing systems (external dilution) as well as separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution). Inherent dilution can e.g. occur after reflux-condenser heat transfer after a small break loss of coolant accident (SBLOCA) with a limited operability of the emergency core cooling (ECC) systems. The TUV Nord e.V. was charged ...

International Science & Technology Center (ISTC)

Investigation of Destruction Mechanisms in Reactor Steels and Alloys under Cycling Deformation

International Nuclear Information System (INIS)

The hardening and embrittlement of reactor pressure vessel (RPV) steels is of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, like vacancies, interstitials, solutes and their clusters. The current procedure to estimate material properties for the irradiated pressure vessels is based on Charpy-V tests of identical material located at the inner shell of the reactor. But the reason for the embrittlement of the materials is not yet totally known. The real nature of the irradiation damage should thus be examined as well as its evolution in time. Fe-Cu binary alloys are often used to mimic the behaviour of such steels. Their study allows. Identifying some of the defects responsible of the hardening, especially when compared to pure iron or C-micro-alloyed iron. More recently the influence of manganese and nickel in low-Cu RPV steels ...

Energy Technology Data Exchange (ETDEWEB)

The growth of stoichiometric and non-stoichiometric silicon nitride films was studied experimentally on 100 mm silicon wafers by batch depositions from the dichlorosilane (SiH{sub 2}Cl{sub 2})-ammonia (NH{sub 3}) system in a hot-wall horizontal low pressure chemical vapor deposition (LPCVD) reactor. The growth kinetics were discussed in terms of the Langmuir adsorption isotherm. The kinetic parameters were determined by comparing the experimental data with a one-dimensional simulation model. The decomposition of NH{sub 3} at high temperatures was included in the simulation procedure. When the SiH{sub 2}Cl{sub 2}:NH{sub 3} ratios were greater than 1.5, a quantity higher than the thermodynamic critical values above which Si-rich nitride films begin to deposit, various SiN{sub x} films with x < 4/3 were obtained. The composition of the SiN{sub x} films was found to vary along the LPCVD reactor. The film stoichiometry was ...

Science.gov (United States)

Pressure Ulcers

International Nuclear Information System (INIS)

This study deals with testing the capability of the code PARET to simulate natural convection cooling phenomena under different boundary conditions. In addition to applying and testing some new options related to simulation of the control rod movement and studying the reactivity effect of thermal expansion fuel elements. The experiments of the simple thermal hydraulic loop of Missouri university about natural cooling phenomena in two narrow paralled channels were used to validate the code. The study indicate good results regarding the distribution of coolant flux velocity and clad temperature. In particular the heat transfer coefficient of natural convection has been calculated in good agreement with the experiment. On the other hand, the core of MNSR reactor has been modelled to simulate the reactor dynamic behaviour under natural convection cooling conditions for different initial power level. The observed oscillation during the initial phase ...

Energy Technology Data Exchange (ETDEWEB)

Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington have focused on developing and evaluating the reliability of nondestructive testing (NDT) approaches for coarse-grained stainless steel reactor components. The objective of this work is to provide information to the United States Nuclear Regulatory Commission (NRC) on the utility, effectiveness and limitation of NDT techniques as related to inservice testing of primary system piping components in pressurized water reactors. We examined cast stainless steel pipe specimens containing thermal and mechanical fatigue cracks located close to the weld roots and having inner and outer diameter surface geometrical conditions that simulate several water reactor primary piping configurations. In addition, segments of vintage centrifugally cast piping were examined to characterize the inherent acoustic noise and scattering ...

Energy Technology Data Exchange (ETDEWEB)

This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approximately}270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented ...