Pickering nuclear fish diversion net

Energy Technology Data Exchange (ETDEWEB)

Xiao, J.; Lew, A. [Ontario Power Generation, Toronto, Ontario (Canada)

2013-07-01

Pickering Fish Diversion Net - An Engineered Environmental Solution that has significantly reduced fish impingement at the Pickering Nuclear Facility. Note: As a recent urgent request/discussed by Mark Elliot, CNE-OPG and Jacques Plourde, CNS.

Pickering unit 1 containment leakage characterization

International Nuclear Information System (INIS)

Zakaib, G.D.

1994-01-01

Results of the design pressure test carried out on Pickering Reactor Building number 1 during late 1992 showed that the leakage rate of the building was close to the safety analysis value of 2.7% contained mass per hour at the design pressure of 41.4 kPa(g) and was significantly higher than that reported after the previous test conducted in the spring of 1987. This unexpected finding initiated the longest and the most comprehensive containment leakage investigation ever undertaken by Ontario Hydro. A thorough investigation of leakage behaviour by repeated testing, inspections, leak search and analysis was launched. The extensive leak search effort included items such as: leak source detection by soap solution application, use of ultrasonic detectors, fogging and tracer gas techniques, systematic systems isolation, thermal imaging of the exterior, and quantification of leak sites by flowmeter and bagging. Using a specially designed volumetric technique, the root cause of the problem was finally confirmed as being due to 'pressure dependent laminar leakage' through the hairline cracks in the dome concrete. Structural analysis indicated that the thermal gradients and pressure loading combined to cause the cracking early in the structure's operating history and that overall structural integrity has not been compromised. Leakage rate analysis using a new fluid mechanics model augmented by the effect of thermal strains indicated that the leakage could be significantly less under certain transient temperature gradient conditions. Several options for repairing the dome were considered by a multidisciplinary team and it was finally decided to apply a specially engineered multilayer elastomeric coating to the exterior concrete surface. When the unit was re-tested in October 1993, a dramatic ten-fold improvement in leakage rate (down to 0.25%/h at design pressure) was observed. This is lower than even the commissioning results and comparable to the performance of newer units

Primary heat transport pump mechanical seal replacement strategy for Pickering B

International Nuclear Information System (INIS)

Chacinsi, V.

1995-01-01

Pickering Nuclear Generating Station is a CANDU PHWR eight unit station located on Lake Ontario. The station is divided into Pickering A (Units 1 to 4) and Pickering B (Units 5 to 8). Pickering B is the focus of this paper. Each unit is rated at 540 MWe. The Primary Heat Transport (PHT) system, which is used to cool the fuel, is divided into four quadrants. Each quadrant has four vertical Byron Jackson PHT main circulation pumps. Three pumps in each quadrant are required for normal operation, leaving one pump in each quadrant as a spare. Each Pickering PHT pump has a Byron Jackson Type SU two stage mechanical seal. The typical pressure breakdown across the seal is 8.7-4.5-1.0 MPa. Certain features of seal operation and the PHT system which influence seal replacement are discussed below. (author)

The 1994 loss of coolant incident at Pickering NGS

Energy Technology Data Exchange (ETDEWEB)

Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

Tuning Amphiphilicity of Particles for Controllable Pickering Emulsion

Directory of Open Access Journals (Sweden)

Zhen Wang

2016-11-01

Full Text Available Pickering emulsions with the use of particles as emulsifiers have been extensively used in scientific research and industrial production due to their edge in biocompatibility and stability compared with traditional emulsions. The control over Pickering emulsion stability and type plays a significant role in these applications. Among the present methods to build controllable Pickering emulsions, tuning the amphiphilicity of particles is comparatively effective and has attracted enormous attention. In this review, we highlight some recent advances in tuning the amphiphilicity of particles for controlling the stability and type of Pickering emulsions. The amphiphilicity of three types of particles including rigid particles, soft particles, and Janus particles are tailored by means of different mechanisms and discussed here in detail. The stabilization-destabilization interconversion and phase inversion of Pickering emulsions have been successfully achieved by changing the surface properties of these particles. This article provides a comprehensive review of controllable Pickering emulsions, which is expected to stimulate inspiration for designing and preparing novel Pickering emulsions, and ultimately directing the preparation of functional materials.

Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

Science.gov (United States)

Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

2013-09-01

Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

Pickering NGS end of commercial operations

Energy Technology Data Exchange (ETDEWEB)

Swami, L. [Ontario Power Generation, Pickering, ON (Canada)

2015-07-01

• Pickering continues as a valued asset for the Province • OPG's current business plan is to continue to operate Pickering to the end of 2020 • Planning its shutdown is underway and includes the following activities to: • Place the units into safe store • Manage the wastes arising • Repurpose the Pickering lands and facilities • Decommissioning will take place in the future and will include appropriate waste management activities. © Copyright 2015 by the Canadian Nuclear Society. All rights reserved.

Nanocellulose-stabilized Pickering emulsions and their applications.

Science.gov (United States)

Fujisawa, Shuji; Togawa, Eiji; Kuroda, Katsushi

2017-01-01

Pickering emulsion, which is an emulsion stabilized by solid particles, offers a wide range of potential applications because it generally provides a more stable system than surfactant-stabilized emulsion. Among various solid stabilizers, nanocellulose may open up new opportunities for future Pickering emulsions owing to its unique nanosizes, amphiphilicity, and other favorable properties (e.g. chemical stability, biodegradability, biocompatibility, and renewability). In this review, the preparation and properties of nanocellulose-stabilized Pickering emulsions are summarized. We also provide future perspectives on their applications, such as drug delivery, food, and composite materials.

Commissioning quality assurance at Pickering NGS

International Nuclear Information System (INIS)

Wieckowski, J.T.

1983-05-01

Ontario Hydro decided in 1978 to implement a formal quality assurance program applicable to commissioning and operation of nuclear generating stations. Pickering NGS is the first station to have the commissioning quality assurance (CQA) program applied to it. This paper outlines the scope, implementation, and evaluation of the CQA program as applied to Pickering Unit 5

Pickering tool management system

International Nuclear Information System (INIS)

Wong, E.H.; Green, A.H.

1997-01-01

Tools were being deployed in the station with no process in effect to ensure that they are maintained in good repair so as to effectively support the performance of Maintenance activities. Today's legal requirements require that all employers have a process in place to ensure that tools are maintained in a safe condition. This is specified in the Ontario Health and Safety Act. The Pickering Tool Management System has been chosen as the process at Pickering N.D to manage tools. Tools are identified by number etching and bar codes. The system is a Windows application installed on several file servers

Seismic analysis of mechanical systems at Pickering NGS

International Nuclear Information System (INIS)

Ghobarah, A.

1995-11-01

The objective of this study is to assess the seismic withstand capacity of selected safety-related mechanical systems associated with the Pressure Relief Duct (PRD) at the Pickering A Nuclear Generating Station. These systems are attached to the PRD and include the Emergency Coolant Injection System piping, the Vacuum Ducts, the Emergency Water Storage System, the PRD expansion joint seals and the PRD to Reactor Building joint seals. The input support motion to the mechanical systems is taken to be the seismic response of the PRD determined in an earlier study using various levels of predetermined ground response spectrum envelope. (author). 12 refs., 13 tabs., 48 figs

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

Rhodes, D.; Metcalfe, R.; Brown, G.

1997-01-01

The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused as least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

Rhodes, D.; Metcalfe, R.; Brown, G.; Kiameh, P.; Burchett, P.

1997-01-01

The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused at least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

Leptospira Exposure and Waste Pickers: A Case-Control Seroprevalence Study in Durango, Mexico

Science.gov (United States)

Alvarado-Esquivel, Cosme; Hernandez-Tinoco, Jesus; Sanchez-Anguiano, Luis Francisco; Ramos-Nevarez, Agar; Cerrillo-Soto, Sandra Margarita; Guido-Arreola, Carlos Alberto

2015-01-01

Background Infection with Leptospira may occur by contact with Leptospira-infected animals. Waste pickers are in contact with rodents and dogs while picking in the garbage. Whether waste pickers are at risk for Leptospira infection is largely unknown. This study was aimed to determine the association of Leptospira IgG seroprevalence with the occupation of waste picking, and to determine the epidemiological characteristics of the waste pickers with Leptospira exposure. Methods Through a case-control study, we determined the seroprevalence of anti-Leptospira IgG antibodies in 90 waste pickers and 90 age- and gender-matched control subjects in Durango City, Mexico using an enzyme immunoassay. Data were analyzed by bivariate and multivariate analyses. Results The prevalence of anti-Leptospira IgG antibodies was similar in waste pickers (4/90: 4.4%) to that in control subjects (5/90: 5.6%) (P = 1.00). Bivariate analysis showed that Leptospira exposure in waste pickers was associated with increasing age (P = 0.009), no education (P = 0.008), and consumption of rat meat (P = 0.04). However, these associations were no longer found by multivariate analysis. Leptospira exposure in waste pickers was not associated with health status, duration in the activity, wearing hand gloves and facemasks, history of injuries with sharp material of the garbage, or contact with animals or soil. Conclusions This is the first study about Leptospira exposure in waste pickers. Results suggest that waste pickers are not at increasing risk for Leptospira exposure in Durango City, Mexico. Further research with a larger sample size to elucidate the association of Leptospira exposure with waste picking activity is needed. PMID:26124911

Leptospira Exposure and Waste Pickers: A Case-Control Seroprevalence Study in Durango, Mexico.

Science.gov (United States)

Alvarado-Esquivel, Cosme; Hernandez-Tinoco, Jesus; Sanchez-Anguiano, Luis Francisco; Ramos-Nevarez, Agar; Cerrillo-Soto, Sandra Margarita; Guido-Arreola, Carlos Alberto

2015-08-01

Infection with Leptospira may occur by contact with Leptospira-infected animals. Waste pickers are in contact with rodents and dogs while picking in the garbage. Whether waste pickers are at risk for Leptospira infection is largely unknown. This study was aimed to determine the association of Leptospira IgG seroprevalence with the occupation of waste picking, and to determine the epidemiological characteristics of the waste pickers with Leptospira exposure. Through a case-control study, we determined the seroprevalence of anti-Leptospira IgG antibodies in 90 waste pickers and 90 age- and gender-matched control subjects in Durango City, Mexico using an enzyme immunoassay. Data were analyzed by bivariate and multivariate analyses. The prevalence of anti-Leptospira IgG antibodies was similar in waste pickers (4/90: 4.4%) to that in control subjects (5/90: 5.6%) (P = 1.00). Bivariate analysis showed that Leptospira exposure in waste pickers was associated with increasing age (P = 0.009), no education (P = 0.008), and consumption of rat meat (P = 0.04). However, these associations were no longer found by multivariate analysis. Leptospira exposure in waste pickers was not associated with health status, duration in the activity, wearing hand gloves and facemasks, history of injuries with sharp material of the garbage, or contact with animals or soil. This is the first study about Leptospira exposure in waste pickers. Results suggest that waste pickers are not at increasing risk for Leptospira exposure in Durango City, Mexico. Further research with a larger sample size to elucidate the association of Leptospira exposure with waste picking activity is needed.

Preparation of Pickering emulsions through interfacial adsorption by soft cyclodextrin nanogels

Directory of Open Access Journals (Sweden)

Shintaro Kawano

2015-11-01

Full Text Available Background: Emulsions stabilized by colloidal particles are known as Pickering emulsions. To date, soft microgel particles as well as inorganic and organic particles have been utilized as Pickering emulsifiers. Although cyclodextrin (CD works as an attractive emulsion stabilizer through the formation of a CD–oil complex at the oil–water interface, a high concentration of CD is normally required. Our research focuses on an effective Pickering emulsifier based on a soft colloidal CD polymer (CD nanogel with a unique surface-active property.Results: CD nanogels were prepared by crosslinking heptakis(2,6-di-O-methyl-β-cyclodextrin with phenyl diisocyanate and subsequent immersion of the resulting polymer in water. A dynamic light scattering study shows that primary CD nanogels with 30–50 nm diameter assemble into larger CD nanogels with 120 nm diameter by an increase in the concentration of CD nanogel from 0.01 to 0.1 wt %. The CD nanogel has a surface-active property at the air–water interface, which reduces the surface tension of water. The CD nanogel works as an effective Pickering emulsion stabilizer even at a low concentration (0.1 wt %, forming stable oil-in-water emulsions through interfacial adsorption by the CD nanogels.Conclusion: Soft CD nanogel particles adsorb at the oil–water interface with an effective coverage by forming a strong interconnected network and form a stable Pickering emulsion. The adsorption property of CD nanogels on the droplet surface has great potential to become new microcapsule building blocks with porous surfaces. These microcapsules may act as stimuli-responsive nanocarriers and nanocontainers.

Pickering education centre aids nuclear acceptance

International Nuclear Information System (INIS)

Anon.

1979-01-01

Activities at the new education centre at Pickering are described. The opening of the Nuclear Communications Centre, in 1978, resulting from a search for an effective means of maintaining public acceptance of Ontario Hydro's extensive nuclear power programme. Activities include participation in the interactive computer games, guided tours of educational exhibits including a model of Pickering A generating station, and displays depicting the Candu fuel cycle, outdoor exhibits of renewable energy sources, and tours of the plant. Outside activities include lectures to schools and citizen, business, or professional groups. (U.K.)

Synthesis of Covalently Cross-Linked Colloidosomes from Peroxidized Pickering Emulsions

Directory of Open Access Journals (Sweden)

Nadiya Popadyuk

2016-10-01

Full Text Available A new approach to the formation of cross-linked colloidosomes was developed on the basis of Pickering emulsions that were stabilized exclusively by peroxidized colloidal particles. Free radical polymerization and a soft template technique were used to convert droplets of a Pickering emulsion into colloidosomes. The peroxidized latex particles were synthesized in the emulsion polymerization process using amphiphilic polyperoxide copolymers poly(2-tert-butylperoxy-2-methyl-5-hexen-3-ine-co-maleic acid (PM-1-MAc or poly[N-(tert-butylperoxymethylacrylamide]-co-maleic acid (PM-2-MAc, which were applied as both initiators and surfactants (inisurfs. The polymerization in the presence of the inisurfs results in latexes with a controllable amount of peroxide and carboxyl groups at the particle surface. Peroxidized polystyrene latex particles with a covalently grafted layer of inisurf PM-1-MAc or PM-2-MAc were used as Pickering stabilizers to form Pickering emulsions. A mixture of styrene and/or butyl acrylate with divinylbenzene and hexadecane was applied as a template for the synthesis of colloidosomes. Peroxidized latex particles located at the interface are involved in the radical reactions of colloidosomes formation. As a result, covalently cross-linked colloidosomes were obtained. It was demonstrated that the structure of the synthesized (using peroxidized latex particles colloidosomes depends on the amount of functional groups and pH during the synthesis. Therefore, the size and morphology of colloidosomes can be controlled by latex particle surface properties.

Data-driven warehouse optimization : Deploying skills of order pickers

NARCIS (Netherlands)

M. Matusiak (Marek); M.B.M. de Koster (René); J. Saarinen (Jari)

2015-01-01

textabstractBatching orders and routing order pickers is a commonly studied problem in many picker-to-parts warehouses. The impact of individual differences in picking skills on performance has received little attention. In this paper, we show that taking into account differences in the skills of

From pioneering to implementing automated blood pressure measurement in clinical practice: Thomas Pickering's legacy

DEFF Research Database (Denmark)

Stolarz-Skrzypek, Katarzyna; Thijs, Lutgarde; Wizner, Barbara

2010-01-01

Thomas G. Pickering spent most of his scientific career in carrying out research on clinical hypertension and blood pressure (BP) measurement. In our review of Pickering's seminal work, we first focused on white-coat hypertension and masked hypertension, two terms that he had introduced. Next, we...... highlighted the early publications of Pickering on diurnal BP variability and on the clinical application of self-measured BP. Pickering's work inspired many investigators worldwide and constituted a solid basis for further research. Pickering's original ideas led to algorithms for risk stratification...

Pickering safeguards: a preliminary analysis

International Nuclear Information System (INIS)

Todd, J.L.; Hodgkinson, J.G.

1977-05-01

A summary is presented of thoughts relative to a systems approach for implementing international safeguards. Included is a preliminary analysis of the Pickering Generating Station followed by a suggested safeguards system for the facility

AECB staff annual report of Pickering NGS for the year 1991

International Nuclear Information System (INIS)

1992-11-01

The AECB Pickering project staff, in cooperation with AECB staff in Ottawa, monitor the operation of Pickering NGS-A units 1-4 and Pickering NGS-B units 5-8 to ensure that Ontario Hydro operates the station in compliance with the licensing and safety requirements of the Atomic Energy Control Board. This report presents the review of licensing issues and station performance during 1991. Improvement over 1990 station operation occurred in the following areas: availability of special safety systems; reduction of the station external dose; reorganization of station management to improve focus; station chemistry; housekeeping and material condition; fuel handling capability; training of operators and maintenance staff. However, little change occurred and improvement is still needed in the following: compliance with operating licence; system surveillance program; station maintenance; environmental qualification; radiation emergency response; fire and rescue emergency response; limited capability to predict and prevent equipment failures such as the boiler tube failure on unit 5. (L.L.)

PROFILE OF PLASTIC WATER BOTTLES WASTES PROCESSING BUSINESS UNIT FOR WASTE PICKERS

Directory of Open Access Journals (Sweden)

Herijanto P.

2017-09-01

Full Text Available Used plastic water bottles waste pickers can be categorized as one of the informal sector’s component. They work for themselves by picking up used water bottles and selling them to the waste collectors. The problem to be solved in this research is How the Most Appropriate Used Plastic Water Bottles Business Model for Waste Pickers Is that enables them to be categorized as formal sector. From the result of the interview with 120 waste pickers, 96 results were qualified to be analyzed. The interview was located in several waste collectors, which were visited by waste pickers at certain hours. The data were analyzed descriptively based on six business aspects. Specifically for production facilities, Quality Function Deployment (QFD and Value Engineering (VE analysis were performed. The results of the analysis indicate that the business is practicable for waste pickers and has the potential to enable them run a formal business sector.

Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

International Nuclear Information System (INIS)

Brown, R.; Cobanoglu, M.M.

1995-01-01

Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

Calculation of homogenized Pickering NGS stainless steel adjuster rod neutron cross-sections using conservation of reaction rates

Energy Technology Data Exchange (ETDEWEB)

Robinson, R C [Atlantic Nuclear Services Ltd. (Canada); Tran, F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

A homogenization methodology for calculation of reactivity device incremental cross-sections has been developed using reaction rate conservation (RRC). A heterogeneous transport calculation of flux was utilised to produce the homogenized cross-sections for a finite difference two group diffusion code. The RRC cross-sections have been shown to improve significantly the prediction of reactivity worth for stainless steel adjuster rods installed in Pickering NGS reactors. (author). 10 refs., 3 tabs., 6 figs.

Water-in-oil Pickering emulsions stabilized by stearoylated microcrystalline cellulose.

Science.gov (United States)

Pang, Bo; Liu, Huan; Liu, Peiwen; Peng, Xinwen; Zhang, Kai

2018-03-01

Hydrophobic particles with static water contact angles larger than 90° are more like to stabilize W/O Pickering emulsions. In particular, high internal phase Pickering emulsions (HIPEs) are of great interest for diverse applications. However, W/O HIPEs have rarely been realized using sustainable biopolymers. Herein, we used stearoylated microcrystalline cellulose (SMCC) to stabilize W/O Pickering emulsions and especially, W/O HIPEs. Moreover, these W/O HIPEs can be further used as platforms for the preparation of porous materials, such as porous foams. Stearoylated microcrystalline cellulose (SMCC) was prepared by modifying MCC with stearoyl chloride under heterogeneous conditions. Using SMCC as emulsifiers, W/O medium and high internal phase Pickering emulsions (MIPEs and HIPEs) with various organic solvents as continuous phases were prepared using one-step and two-step methods, respectively. Polystyrene (PS) foams were prepared after polymerization of oil phase using HIPEs as templates and their oil/water separation capacity were studied. SMCC could efficiently stabilize W/O Pickering emulsions and HIPEs could only be prepared via the two-step method. The internal phase volume fraction of the SMCC-stabilized HIPEs reached as high as 89%. Diverse internal phase volume fractions led to distinct inner structures of foams with closed or open cells. These macroporous polystyrene (PS) foams demonstrated great potential for the effective absorption of organic solvents from underwater. Copyright © 2017 Elsevier Inc. All rights reserved.

Pickering seismic safety margin

International Nuclear Information System (INIS)

Ghobarah, A.; Heidebrecht, A.C.; Tso, W.K.

1992-06-01

A study was conducted to recommend a methodology for the seismic safety margin review of existing Canadian CANDU nuclear generating stations such as Pickering A. The purpose of the seismic safety margin review is to determine whether the nuclear plant has sufficient seismic safety margin over its design basis to assure plant safety. In this review process, it is possible to identify the weak links which might limit the seismic performance of critical structures, systems and components. The proposed methodology is a modification the EPRI (Electric Power Research Institute) approach. The methodology includes: the characterization of the site margin earthquake, the definition of the performance criteria for the elements of a success path, and the determination of the seismic withstand capacity. It is proposed that the margin earthquake be established on the basis of using historical records and the regional seismo-tectonic and site specific evaluations. The ability of the components and systems to withstand the margin earthquake is determined by database comparisons, inspection, analysis or testing. An implementation plan for the application of the methodology to the Pickering A NGS is prepared

Evolution of CANDU reactor design

International Nuclear Information System (INIS)

Pon, G.A.

1978-08-01

The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

Fuel deposits, chemistry and CANDU® reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2014-01-01

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

Estimating the possible range of recycling rates achieved by dump waste pickers: The case of Bantar Gebang in Indonesia.

Science.gov (United States)

Sasaki, Shunsuke; Araki, Tetsuya

2014-06-01

This article presents informal recycling contributions made by scavengers in the surrounding area of Bantar Gebang final disposal site for municipal solid waste generated in Jakarta. Preliminary fieldwork was conducted through daily conversations with scavengers to identify recycling actors at the site, and then quantitative field surveys were conducted twice. The first survey (n = 504 households) covered 33% of all households in the area, and the second survey (n = 69 households) was conducted to quantify transactions of recyclables among scavengers. Mathematical equations were formulated with assumptions made to estimate the possible range of recycling rates achieved by dump waste pickers. Slightly over 60% of all respondents were involved in informal recycling and over 80% of heads of households were waste pickers, normally referred to as live-in waste pickers and live-out waste pickers at the site. The largest percentage of their spouses were family workers, followed by waste pickers and housewives. Over 95% of all households of respondents had at least one waste picker or one small boss who has a coequal status of a waste picker. Average weight of recyclables collected by waste pickers at the site was estimated to be approximately 100 kg day(-1) per household on the net weight basis. The recycling rate of solid wastes collected by all scavengers at the site was estimated to be in the range of 2.8-7.5% of all solid wastes transported to the site. © The Author(s) 2014.

Moderator inlet line hanger replacement for Pickering nuclear power station

International Nuclear Information System (INIS)

Kirkpatrick, R.A.; Bowman, J.M.; Symmons, W.R.; El-Nesr, S.

1988-01-01

Ontario Hydro's Pickering Nuclear Generating Station (PNGS), Units 1 and 2 were shutdown for large scale fuel channel replacement. Other nonroutine inspection and maintenance activities were performed to determine the overall condition of the units and it was seen that a moderator inlet line hanger (identified as HR-29) had failed in both units. Subsequent inspections during planned maintenance outages of Pickering NGS Units 3 and 4 revealed that hanger HR-29 had failed and required replacement. A research program was conducted to find a suitable technique. These problems included accessing tooling through small inspection ports, manipulating tooling from a significant distance and the high radiation fields within the vault. This paper describes the program undertaken to replace hanger HR-29. (author)

Edible foam based on Pickering effect of probiotic bacteria and milk proteins

DEFF Research Database (Denmark)

Yücel, Cigdem; Geng, Xiaolu; Cárdenas, Marité

2017-01-01

We report the preparation and characterization of aqueous Pickering foams using bio-particles constituted by lactic acid bacteria surface modified by oppositely charged milk proteins. Cell surface modification was shown by zeta potential measurements. Foams stabilized by bacterial Pickering bio-p...

Superhydrophobic cellulose-based bionanocomposite films from Pickering emulsions

Science.gov (United States)

Bayer, Ilker S.; Steele, Adam; Martorana, Philip J.; Loth, Eric; Miller, Lance

2009-04-01

Inherently superhydrophobic and flexible cellulose-based bionanocomposites were fabricated from solid stabilized (Pickering) emulsions. Emulsions were formed by dispersing cyclosiloxanes in water stabilized by layered silicate particles and were subsequently modified by blending into a zinc oxide nanofluid. The polymer matrix was a blend of cellulose nitrate and fluoroacrylic polymer (Zonyl 8740) precompatibilized in solution. Coatings were spray cast onto aluminum substrates from polymer blends dispersed in modified Pickering emulsions. No postsurface treatment was required to induce superhydrophobicity. Effect of antiseptic additives on bionanocomposite superhydrophobicity is also discussed. Replacing cellulose nitrate with commercial liquid bandage solutions produced identical superhydrophobic coatings.

Ergonomic Evaluation of Battery Powered Portable Cotton Picker

Science.gov (United States)

Dixit, A.; Manes, G. S.; Singh, A.; Prakash, A.; Mahal, J. S.

2012-09-01

Ergonomic evaluation of battery powered portable manual cotton picker was carried out on two subjects for three cotton varieties and was compared against manual method of picking. It is a hand operated machine and has a pair of chain with small sharp edged teeth and sprockets and is operated by a light weight 12 V battery. Cotton gets entangled with the chain and is collected and guided into the collection bag. Average heart rate, oxygen consumption, workload, energy expenditure was more in case of cotton picking by manual cotton picker as compared to manual picking for both the subjects for all three cotton variety types. Oxygen consumption varied from 0.81 to 0.97 l/min, workload varied from 36.32 to 46.16 W and energy expenditure varied from 16.83 to 20.33 kJ/min for both the subject in case of machine picking for all three cotton varieties. The maximum discomfort experienced by the subjects during picking cotton by manual cotton picker was in right wrist palm, right forearm, upper and lower back, left shoulder and in lower legs and both feet.

Novel carboxymethyl cellulose-polyvinyl alcohol blend films stabilized by Pickering emulsion incorporation method.

Science.gov (United States)

Fasihi, Hadi; Fazilati, Mohammad; Hashemi, Mahdi; Noshirvani, Nooshin

2017-07-01

The aim of this study was to investigate the possibility of increasing the antimicrobial and antioxidant properties of biodegradable active films stabilized via Pickering emulsions. The blend films were prepared from carboxymethyl cellulose (CMC) and polyvinyl alcohol (PVA), emulsified with oleic acid (OL) and incorporated with rosemary essential oil (REO). Formation of Pickering emulsion was confirmed by scanning electron microscopy (SEM), optical microscopy, mean droplet size and emulsion stability. Morphological, optical, physical, mechanical, thermal, antifungal and antioxidant properties of the films incorporated with different concentrations of REO (0.5, 1.5 and 3%) were determined. The results showed an increase in UV absorbance and elongation at break but, a decrease in tensile strength and thermal stability of the films. Interestingly, films containing REO exhibited considerable antioxidant and antimicrobial properties. In vitro microbial tests exhibited 100% fungal inhibition against Penicillium digitatum in the films containing 3% REO. In addition, no fungal growth were observed after 60days of storage at 25°C in bread slices were stored with active films incorporated with 3% REO, could attributed to the slow and regular release of REO caused by Pickering emulsions. The results of this study suggest that Pickering emulsion is a very promising method, which significantly affects antioxidant and antimicrobial activities of the films. Copyright © 2017 Elsevier Ltd. All rights reserved.

Wet steam turbines for CANDU-Reactors

International Nuclear Information System (INIS)

Westmacott, C.H.L.

1977-01-01

The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

The Pickering mesonet 1988 data report

International Nuclear Information System (INIS)

Salmon, J.R.; Taylor, P.A.

1989-10-01

This report describes the demonstration mesoscale meteorological monitoring network (mesonet) installed in the vicinity of the Pickering Nuclear Generating Station. It also summarizes the data collected by the network during 1988 and provides some examples of situations in which mesoscale effects dominate the local wind flow

Safe "cloudification" of large images through picker APIs.

Science.gov (United States)

Bremer, Erich; Kurc, Tahsin; Gao, Yi; Saltz, Joel; Almeida, Jonas S

2016-01-01

The "Box model" allows users with no particular training in informatics, or access to specialized infrastructure, operate generic cloud computing resources through a temporary URI dereferencing mechanism known as "drop-file-picker API" ("picker API" for sort). This application programming interface (API) was popularized in the web app development community by DropBox, and is now a consumer-facing feature of all major cloud computing platforms such as Box.com, Google Drive and Amazon S3. This reports describes a prototype web service application that uses picker APIs to expose a new, "cloudified", API tailored for image analysis, without compromising the private governance of the data exposed. In order to better understand this cross-platform cloud computing landscape, we first measured the time for both transfer and traversing of large image files generated by whole slide imaging (WSI) in Digital Pathology. The verification that there is extensive interconnectivity between cloud resources let to the development of a prototype software application that exposes an image-traversing REST API to image files stored in any of the consumer-facing "boxes". In summary, an image file can be upload/synchronized into a any cloud resource with a file picker API and the prototype service described here will expose an HTTP REST API that remains within the safety of the user's own governance. The open source prototype is publicly available at sbu-bmi.github.io/imagebox. Availability The accompanying prototype application is made publicly available, fully functional, with open source, at http://sbu-bmi.github.io/imagebox://sbu-bmi.github.io/imagebox. An illustrative webcasted use of this Web App is included with the project codebase at https://github.com/SBU-BMI/imageboxs://github.com/SBU-BMI/imagebox.

DeepPicker: A deep learning approach for fully automated particle picking in cryo-EM.

Science.gov (United States)

Wang, Feng; Gong, Huichao; Liu, Gaochao; Li, Meijing; Yan, Chuangye; Xia, Tian; Li, Xueming; Zeng, Jianyang

2016-09-01

Particle picking is a time-consuming step in single-particle analysis and often requires significant interventions from users, which has become a bottleneck for future automated electron cryo-microscopy (cryo-EM). Here we report a deep learning framework, called DeepPicker, to address this problem and fill the current gaps toward a fully automated cryo-EM pipeline. DeepPicker employs a novel cross-molecule training strategy to capture common features of particles from previously-analyzed micrographs, and thus does not require any human intervention during particle picking. Tests on the recently-published cryo-EM data of three complexes have demonstrated that our deep learning based scheme can successfully accomplish the human-level particle picking process and identify a sufficient number of particles that are comparable to those picked manually by human experts. These results indicate that DeepPicker can provide a practically useful tool to significantly reduce the time and manual effort spent in single-particle analysis and thus greatly facilitate high-resolution cryo-EM structure determination. DeepPicker is released as an open-source program, which can be downloaded from https://github.com/nejyeah/DeepPicker-python. Copyright © 2016 Elsevier Inc. All rights reserved.

Pickering interfacial catalysis for biphasic systems: from emulsion design to green reactions.

Science.gov (United States)

Pera-Titus, Marc; Leclercq, Loïc; Clacens, Jean-Marc; De Campo, Floryan; Nardello-Rataj, Véronique

2015-02-09

Pickering emulsions are surfactant-free dispersions of two immiscible fluids that are kinetically stabilized by colloidal particles. For ecological reasons, these systems have undergone a resurgence of interest to mitigate the use of synthetic surfactants and solvents. Moreover, the use of colloidal particles as stabilizers provides emulsions with original properties compared to surfactant-stabilized emulsions, microemulsions, and micellar systems. Despite these specific advantages, the application of Pickering emulsions to catalysis has been rarely explored. This Minireview describes very recent examples of hybrid and composite amphiphilic materials for the design of interfacial catalysts in Pickering emulsions with special emphasis on their assets and challenges for industrially relevant biphasic reactions in fine chemistry, biofuel upgrading, and depollution. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

The One-Step Pickering Emulsion Polymerization Route for Synthesizing Organic-Inorganic Nanocomposite Particles

Directory of Open Access Journals (Sweden)

Kaushal Rege

2010-02-01

Full Text Available Polystyrene-silica core-shell nanocomposite particles are successfully prepared via one-step Pickering emulsion polymerization. Possible mechanisms of Pickering emulsion polymerization are addressed in the synthesis of polystyrene-silica nanocomposite particles using 2,2-azobis(2-methyl-N-(2-hydroxyethylpropionamide (VA-086 and potassium persulfate (KPS as the initiator. Motivated by potential applications of “smart” composite particles in controlled drug delivery, the one-step Pickering emulsion polymerization route is further applied to synthesize polystyrene/poly(N-isopropylacrylamide (PNIPAAm-silica core-shell nanoparticles with N-isopropylacrylamide incorporated into the core as a co-monomer. The polystyrene/PNIPAAm-silica composite nanoparticles are temperature sensitive and can be taken up by human prostate cancer (PC3-PSMA cells.

From pioneering to implementing automated blood pressure measurement in clinical practice: Thomas Pickering's legacy

DEFF Research Database (Denmark)

Stolarz-Skrzypek, Katarzyna; Thijs, Lutgarde; Wizner, Barbara

2010-01-01

Thomas G. Pickering spent most of his scientific career in carrying out research on clinical hypertension and blood pressure (BP) measurement. In our review of Pickering's seminal work, we first focused on white-coat hypertension and masked hypertension, two terms that he had introduced. Next, we...

Women cotton pickers perceptions about health hazards due to pesticide use in irrigated punjab

International Nuclear Information System (INIS)

Abbas, M.; Mehmood, I.; Bashir, A.; Hassan, S.

2015-01-01

In Pakistan, cotton crop has special importance from the perspective of largest employment generation both for males and females in the production and value chains. Cotton picking is primarily a female specific activity in all cropping zones of Pakistan. Women cotton pickers mostly belong to poor rural society involved in this labour force to feed their families. Cotton pickers in Pakistan face some serious health related problems due to heavy use of pesticides on cotton crop. The present study was designed to investigate the problem faced by women cotton pickers and their role in household decision making. Overall 150 women cotton pickers were interviewed from Bahawalnagar, Sahiwal and Vehari districts of cotton-wheat zone of the Punjab. Summary statistics of women cotton pickers' showed mean average age was 33 years and had 2.4 ears of formal schooling and 10 years of cotton picking experience. The main reasons for cotton picking reported were to reduce family financial burden (30%) followed by better access to food and resource (23%) and better education of children (21%). Majority of the respondents (97.33%) reported that the mode of payments of cotton picking was in cash and the most of the respondents (83.70%) reported that they got wages in time. Only few respondents (8.70%) were aware of health hazards due to pesticides and only 10% women wear protective clothes during cotton picking. Majority of the respondents (76%) wash their clothes after cotton picking whereas almost all the respondents wash their hand after cotton picking. The women cotton pickers faced health problem, tiredness (54.5%), mental disturbance (9.90%) and fatigue (8.00%). More than 58% women reported their involvement in household decision making regarding food and groceries while 30.6% women involved in decision about education of children. It is suggested that the female cotton pickers should be educated about the importance (in terms of disease treatment and long-run health costs

Safe “cloudification” of large images through picker APIs

Science.gov (United States)

Bremer, Erich; Kurc, Tahsin; Gao, Yi; Saltz, Joel; Almeida, Jonas S

2016-01-01

The “Box model” allows users with no particular training in informatics, or access to specialized infrastructure, operate generic cloud computing resources through a temporary URI dereferencing mechanism known as “drop-file-picker API” (“picker API” for sort). This application programming interface (API) was popularized in the web app development community by DropBox, and is now a consumer-facing feature of all major cloud computing platforms such as Box.com, Google Drive and Amazon S3. This reports describes a prototype web service application that uses picker APIs to expose a new, “cloudified”, API tailored for image analysis, without compromising the private governance of the data exposed. In order to better understand this cross-platform cloud computing landscape, we first measured the time for both transfer and traversing of large image files generated by whole slide imaging (WSI) in Digital Pathology. The verification that there is extensive interconnectivity between cloud resources let to the development of a prototype software application that exposes an image-traversing REST API to image files stored in any of the consumer-facing “boxes”. In summary, an image file can be upload/synchronized into a any cloud resource with a file picker API and the prototype service described here will expose an HTTP REST API that remains within the safety of the user’s own governance. The open source prototype is publicly available at sbu-bmi.github.io/imagebox. Availability The accompanying prototype application is made publicly available, fully functional, with open source, at http://sbu-bmi.github.io/imagebox://sbu-bmi.github.io/imagebox. An illustrative webcasted use of this Web App is included with the project codebase at https://github.com/SBU-BMI/imageboxs://github.com/SBU-BMI/imagebox. PMID:28269829

Fuel deposits, chemistry and CANDU reactor operation

International Nuclear Information System (INIS)

Roberts, J.G.

2013-01-01

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

Cobalt-60 control in Ontario Hydro reactors

International Nuclear Information System (INIS)

Lacy, C.S.

1988-01-01

This paper discusses the impact of specifying reduced Cobalt-59 in the primary heat transport circuit materials of construction on the radiation fields developed around the primary circuit. An eight-fold reduction in steam generator radiation fields due to Cobalt-60 has been observed for two identical sets of reactors, one with and one without Cobalt-59 control. The comparison is between eight reactors at the Pickering Nuclear Generating Station (PNGS). Units 5 to 8 (PNGS-B) are identical to Units 1 to 4 (PNGS-A) except that PNGS-B has reduced impurity Cobalt-59 in the alloys of construction and a reduced use of stellite. The effects of chemistry control are also discussed

Analytical Diagnostics of Non-Optimal Use of Pesticides and Health Hazards for Vegetable Pickers

International Nuclear Information System (INIS)

Zafar, M.; Mehmood, T.; Baig, I. A.; Saboor, A.; Sadiq, S.; Mahmood, K.

2016-01-01

Economically pesticides are meant to control pests in the fields. Up to certain optimal use of a typical pesticide, it enhances the yield of crops and vegetables. But, eventually amplified use of pesticides results in contamination of environment (water, soil, and air) and increase the health cost of vegetable pickers. The purpose of this study is to estimate the excessive use of pesticides and economic cost of health hazards for the vegetable pickers in district Vehari. Data from 90 respondents were collected and analyzed. The most common health problems identified during the survey were headache, eye irritation, skin infection, cough and shortness of breath. Health cost consists of costs related to precautionary measure, medication, traveling, the opportunity cost of attended persons and productivity loss. The mean health cost of vegetable pickers in the study area was about Rs. 385 per picker per year. Health cost model was used to measure the health cost of vegetable pickers. The regression results showed that pesticides were being applied non-optimally in the study area i.e., number of pesticide applications for vegetables (7-31) were substantially higher than the recommended dose. Health cost function was significantly different from zero as indicated by F-stat (32.18) and it is also supported by R/sup 2/ that about 70 percent variation in health cost is explained by medication accompanied by productivity loss (Rs. 223), precautionary measure (Rs. 134), attended person cost (Rs. 14) and traveling expenditures (Rs. 16). Hence, strict legislation is required to overcome the availability of hazardous pesticides and to keep the vegetable pickers aware of the optimal use of pesticides through appropriate extension services. (author)

Field testing of behavioral barriers for cooling water intake structures -test site 1 - Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Patrick, P.H.; McKinley, R.S.; Micheletti, W.C.

1988-01-01

A multi-year research program was developed by the Electric Power Research Institute to evaluate the effectiveness of selected behavioral systems for fish exclusion at sites representative of different aquatic environments. The first test site was the Pickering Nuclear Generating Station (NGS) located on Lake Ontario which represented the Great Lakes environment. A single pneumatic popper, a low frequency, high amplitude sound deterrent, was found to effectively exclude adult alewife, the principal species impinged at Pickering NGS. An air bubble curtain, used either alone or combined with strobe lights, was not a consistent deterrent. Effectiveness of air bubbles was only enhanced when used in association with a popper. Strobe lights were the least effective of the three devices tested. Operation of all three devices together did not surpass the effectiveness of the popper when used alone. Sound deterrents show promise for fish exclusion at generating stations located on the Great Lakes

Warehouse order-picking process. Order-picker routing problem

Directory of Open Access Journals (Sweden)

E. V. Korobkov

2015-01-01

Full Text Available This article continues “Warehouse order-picking process” cycle and describes order-picker routing sub-problem of a warehouse order-picking process. It draws analogies between the orderpickers’ routing problem and traveling salesman’s problem, shows differences between the standard problem statement of a traveling salesman and routing problem of warehouse orderpickers, and gives the particular Steiner’s problem statement of a traveling salesman.Warehouse layout with a typical order is represented by a graph, with some its vertices corresponding to mandatory order-picker’s visits and some other ones being noncompulsory. The paper describes an optimal Ratliff-Rosenthal algorithm to solve order-picker’s routing problem for the single-block warehouses, i.e. warehouses with only two crossing aisles, defines seven equivalent classes of partial routing sub-graphs and five transitions used to have an optimal routing sub-graph of a order-picker. An extension of optimal Ratliff-Rosenthal order-picker routing algorithm for multi-block warehouses is presented and also reasons for using the routing heuristics instead of exact optimal algorithms are given. The paper offers algorithmic description of the following seven routing heuristics: S-shaped, return, midpoint, largest gap, aisle-by-aisle, composite, and combined as well as modification of combined heuristics. The comparison of orderpicker routing heuristics for one- and two-block warehouses is to be described in the next article of the “Warehouse order-picking process” cycle.

Development and characterization of novel antimicrobial bilayer films based on Polylactic acid (PLA)/Pickering emulsions.

Science.gov (United States)

Zhu, Jun-You; Tang, Chuan-He; Yin, Shou-Wei; Yang, Xiao-Quan

2018-02-01

Biodegradable food packaging is sustainable and has a great application prospect. PLA is a promising alternative for petroleum-derived polymers. However, PLA packaging suffers from poor barrier properties compared with petroleum-derived ones. To address this issue, we designed bilayer films based on PLA and Pickering emulsions. The formed bilayer films were compact and uniform and double layers were combined firmly. This strategy enhanced mechanical resistance, ductility and moisture barrier of Pickering emulsion films, and concomitantly enhanced the oxygen barrier for PLA films. Thymol loadings in Pickering emulsion layer endowed them with antimicrobial and antioxidant activity. The release profile of thymol was well fitted with Fick's second law. The antimicrobial activity of the films depended on film types, and Pickering emulsion layer presented larger inhibition zone than PLA layer, hinting that the films possessed directional releasing role. This study opens a promising route to fabricate bilayer architecture creating synergism of each layer. Copyright © 2017 Elsevier Ltd. All rights reserved.

Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

International Nuclear Information System (INIS)

Serdula, K.J.

1988-01-01

Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

Pickering emulsions for skin decontamination.

Science.gov (United States)

Salerno, Alicia; Bolzinger, Marie-Alexandrine; Rolland, Pauline; Chevalier, Yves; Josse, Denis; Briançon, Stéphanie

2016-08-01

This study aimed at developing innovative systems for skin decontamination. Pickering emulsions, i.e. solid-stabilized emulsions, containing silica (S-PE) or Fuller's earth (FE-PE) were formulated. Their efficiency for skin decontamination was evaluated, in vitro, 45min after an exposure to VX, one of the most highly toxic chemical warfare agents. Pickering emulsions were compared to FE (FE-W) and silica (S-W) aqueous suspensions. PE containing an oil with a similar hydrophobicity to VX should promote its extraction. All the formulations reduced significantly the amount of VX quantified on and into the skin compared to the control. Wiping the skin surface with a pad already allowed removing more than half of VX. FE-W was the less efficient (85% of VX removed). The other formulations (FE-PE, S-PE and S-W) resulted in more than 90% of the quantity of VX removed. The charge of particles was the most influential factor. The low pH of formulations containing silica favored electrostatic interactions of VX with particles explaining the better elimination from the skin surface. Formulations containing FE had basic pH, and weak interactions with VX did not improve the skin decontamination. However, these low interactions between VX and FE promote the transfer of VX into the oil droplets in the FE-PE. Copyright © 2016 Elsevier B.V. All rights reserved.

The number of pickers and stock-keeping unit arrangement on a uni-directional picking line

Directory of Open Access Journals (Sweden)

Hagspihl, Robert

2014-10-01

Full Text Available The order picking process is often the single largest expense in a distribution centre (DC. The DC considered in this paper uses a picking line configuration to perform order picking. The number of pickers in a picking line, and the initial arrangement of stock-keeping units (SKUs, are two important factors that affect the total completion time of the picking lines. In this paper, the picking line configuration is simulated with an agent-based approach to describe the behaviour of an individual picker. The simulation is then used to analyse the effect of the number of pickers and the SKU arrangement. Verification and validation of this model shows that the model represents the real-world picking line to a satisfactory degree. Marginal analysis (MA was chosen to determine a ‘good’ number of pickers by means of the simulation model. A look-up table is presented to provide decision support for the choice of a ‘good’ number of pickers to improve completion times of the picking line, for the properties of a specific picking line. The initial SKU arrangement on a picking line is shown to be a factor that can affect the level of picker congestion and the total completion time. The greedy ranking and partitioning (GRP and organ pipe arrangement (OPA techniques from the literature, as well as the historical SKU arrangements used by the retailer under consideration, were compared with the proposed classroom discipline heuristic (CDH for SKU arrangement. It was found that the CDH provides an more even spread of SKUs that are picked most frequently, thus decreasing congestion and total completion time.

Exploiting the pliability and lateral mobility of Pickering emulsion for enhanced vaccination

Science.gov (United States)

Xia, Yufei; Wu, Jie; Wei, Wei; Du, Yiqun; Wan, Tao; Ma, Xiaowei; An, Wenqi; Guo, Aiying; Miao, Chunyu; Yue, Hua; Li, Shuoguo; Cao, Xuetao; Su, Zhiguo; Ma, Guanghui

2018-02-01

A major challenge in vaccine formulations is the stimulation of both the humoral and cellular immune response for well-defined antigens with high efficacy and safety. Adjuvant research has focused on developing particulate carriers to model the sizes, shapes and compositions of microbes or diseased cells, but not antigen fluidity and pliability. Here, we develop Pickering emulsions--that is, particle-stabilized emulsions that retain the force-dependent deformability and lateral mobility of presented antigens while displaying high biosafety and antigen-loading capabilities. Compared with solid particles and conventional surfactant-stabilized emulsions, the optimized Pickering emulsions enhance the recruitment, antigen uptake and activation of antigen-presenting cells, potently stimulating both humoral and cellular adaptive responses, and thus increasing the survival of mice upon lethal challenge. The pliability and lateral mobility of antigen-loaded Pickering emulsions may provide a facile, effective, safe and broadly applicable strategy to enhance adaptive immunity against infections and diseases.

Bruce used fuel dry storage project evolution from Pickering to Bruce

International Nuclear Information System (INIS)

Young, R.E.

1996-01-01

Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container's capacity increased to 600 bundles; the container's lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs

Bruce used fuel dry storage project evolution from Pickering to Bruce

Energy Technology Data Exchange (ETDEWEB)

Young, R E [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1997-12-31

Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container`s capacity increased to 600 bundles; the container`s lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs.

Heterogeneous Pd catalysts as emulsifiers in Pickering emulsions for integrated multistep synthesis in flow chemistry.

Science.gov (United States)

Hiebler, Katharina; Lichtenegger, Georg J; Maier, Manuel C; Park, Eun Sung; Gonzales-Groom, Renie; Binks, Bernard P; Gruber-Woelfler, Heidrun

2018-01-01

Within the "compartmentalised smart factory" approach of the ONE-FLOW project the implementation of different catalysts in "compartments" provided by Pickering emulsions and their application in continuous flow is targeted. We present here the development of heterogeneous Pd catalysts that are ready to be used in combination with biocatalysts for catalytic cascade synthesis of active pharmaceutical ingredients (APIs). In particular, we focus on the application of the catalytic systems for Suzuki-Miyaura cross-coupling reactions, which is the key step in the synthesis of the targeted APIs valsartan and sacubitril. An immobilised enzyme will accomplish the final product formation via hydrolysis. In order to create a large interfacial area for the catalytic reactions and to keep the reagents separated until required, the catalyst particles are used to stabilise Pickering emulsions of oil and water. A set of Ce-Sn-Pd oxides with the molecular formula Ce 0.99- x Sn x Pd 0.01 O 2-δ ( x = 0-0.99) has been prepared utilising a simple single-step solution combustion method. The high applicability of the catalysts for different functional groups and their minimal leaching behaviour is demonstrated with various Suzuki-Miyaura cross-coupling reactions in batch as well as in continuous flow employing the so-called "plug & play reactor". Finally, we demonstrate the use of these particles as the sole emulsifier of oil-water emulsions for a range of oils.

Pickering Emulsions for Food Applications: Background, Trends, and Challenges

NARCIS (Netherlands)

Berton-Carabin, C.C.; Schroën, C.G.P.H.

2015-01-01

Particle-stabilized emulsions, also referred to as Pickering emulsions, have garnered exponentially increasing interest in recent years. This has also led to the first food applications, although the number of related publications is still rather low. The involved stabilization mechanisms are

Tailoring the Wettability of Colloidal Particles for Pickering Emulsions via Surface Modification and Roughness

Directory of Open Access Journals (Sweden)

Meina Xiao

2018-06-01

Full Text Available Pickering emulsions are water or oil droplets that are stabilized by colloidal particles and have been intensely studied since the late 90s. The surfactant-free nature of these emulsions has little adverse effects such as irritancy and contamination of environment and typically exhibit enhanced stability compared to surfactant-stabilized emulsions. Therefore, they offer promising applications in cosmetics, food science, controlled release, and the manufacturing of microcapsules and porous materials. The wettability of the colloidal particles is the main parameter determining the formation and stability of Pickering emulsions. Tailoring the wettability by surface chemistry or surface roughness offers considerable scope for the design of a variety of hybrid nanoparticles that may serve as novel efficient Pickering emulsion stabilizers. In this review, we will discuss the recent advances in the development of surface modification of nanoparticles.

Particle Shape Anisotropy in Pickering Emulsions: Cubes and Peanuts

NARCIS (Netherlands)

de Folter, J.W.J.; Hutter, E.M.; Castillo, S.I.R.; Klop, K.E.; Philipse, A.P.; Kegel, W.K.

2014-01-01

We have investigated the effect of particle shape in Pickering emulsions by employing, for the first time, cubic and peanut-shaped particles. The interfacial packing and orientation of anisotropic microparticles are revealed at the single-particle level by direct microscopy observations. The uniform

Pickering emulsion: A novel template for microencapsulated phase change materials with polymer–silica hybrid shell

International Nuclear Information System (INIS)

Yin, Dezhong; Ma, Li; Liu, Jinjie; Zhang, Qiuyu

2014-01-01

MePCMs (microencapsulated phase change materials) with covalently bonded SiO 2 /polymer hybrid as shell were fabricated via Pickering emulsion polymerization stabilized solely by organically-modified SiO 2 particles. Morphology and core–shell structure of these microcapsules were observed by scanning electron microscopy (SEM). Thermal properties of microencapsulated 1-dodecanol were determined using DSC (differential scanning calorimetry) and TGA (thermal gravimetric analysis). The results indicate that mass ratio of St (styrene)/DVB (divinylbenzene)/dodecanol has great effect on the morphology, inner structure, microencapsulation efficiency and durability of resultant MePCMs. When ratio of St/DVB/dodecanol was 5/1/12, dodecanol content of as much as 62.8% is obtained and the utility efficiency of dodecanol reaches 94.2%. The prepared MePCMs present good durability and thermal reliability. 2.2% of core material leached away the microcapsule after suspended in water for 10 days and 5.8% of core material leached after 2000 accelerated thermal cycling. Our study demonstrated that Pickering emulsion polymerization is a simple and robust method for the preparation of MePCMs with polymer–inorganic hybrids as shell. - Highlights: • We fabricated MePCM via surfactant-free Pickering emulsion polymerization. • The shell of MePCM was composed of PS/SiO 2 organic–inorganic hybrids. • The phase change enthalpy of MePCM is 125.0 J g −1 and the utility efficiency of 1-dodecanol reached 94.2%. • Only 2.2% and 5.8% of core material lost after durability test and 2000 accelerated thermal cycling respectively

Seismic assessment of the Pickering pressure relief duct

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objectives of the study are to examine the structural response of the Pickering pressure relief duct when subjected to earthquake ground motion and to estimate the seismic withstand capacity of various components of the structural system on the basis of performance criteria consistent with the safety function of the duct. (author). 24 refs., 16 tabs., 31 figs

Aquivion Perfluorosulfonic Superacid as an Efficient Pickering Interfacial Catalyst for the Hydrolysis of Triglycerides.

Science.gov (United States)

Shi, Hui; Fan, Zhaoyu; Hong, Bing; Pera-Titus, Marc

2017-09-11

Rational design of the surface properties of heterogeneous catalysts can boost the interfacial activity in biphasic reactions through the generation of Pickering emulsions. This concept, termed Pickering interfacial catalysis (PIC), has shown promising credentials in acid-catalyzed transesterification, ester hydrolysis, acetalization, etherification, and alkylation reactions. PIC has now been applied to the efficient, solvent-free hydrolysis of the triglyceride glyceryl trilaurate to lauric acid, catalyzed by Aquivion perfluorosulfonic superacid at mild conditions (100 °C and ambient pressure). © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Study of the time effect on the strength of cell-cell adhesion force by a novel nano-picker

Energy Technology Data Exchange (ETDEWEB)

Shen, Yajing, E-mail: shen@robo.mein.naogya-u.ac.jp [Dept. of Micro-Nano Systems Engineering, Nagoya University, Nagoya 464-8603 (Japan); Nakajima, Masahiro [Center for Micro-Nano Mechatronics, Nagoya University, Nagoya 464-8603 (Japan); Kojima, Seiji; Homma, Michio [Division of Biological Science, Nagoya University, Nagoya 464-8603 (Japan); Fukuda, Toshio [Dept. of Micro-Nano Systems Engineering, Nagoya University, Nagoya 464-8603 (Japan); Center for Micro-Nano Mechatronics, Nagoya University, Nagoya 464-8603 (Japan)

2011-06-03

Highlights: {yields} A nano-picker is developed for single cell adhesion force measurement. {yields} The adhesion of picker-cell has no influence to the cell-cell measurement result. {yields} Cell-cell adhesion force has a rise at the first few minutes and then becomes constant. -- Abstract: Cell's adhesion is important to cell's interaction and activates. In this paper, a novel method for cell-cell adhesion force measurement was proposed by using a nano-picker. The effect of the contact time on the cell-cell adhesion force was studied. The nano-picker was fabricated from an atomic force microscopy (AFM) cantilever by nano fabrication technique. The cell-cell adhesion force was measured based on the deflection of the nano-picker beam. The result suggests that the adhesion force between cells increased with the increasing of contact time at the first few minutes. After that, the force became constant. This measurement methodology was based on the nanorobotic manipulation system inside an environmental scanning electron microscope. It can realize both the observation and manipulation of a single cell at nanoscale. The quantitative and precise cell-cell adhesion force result can be obtained by this method. It would help us to understand the single cell interaction with time and would benefit the research in medical and biological fields potentially.

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Condition based maintenance pilot projects at Pickering ND

International Nuclear Information System (INIS)

Zemdegs, R.T.

1995-01-01

Ontario Hydro has recognized that the approaches to maintenance have undergone significant changes to the past decades. The traditional break down maintenance approach has been replaced by preventative maintenance and more recently, by condition based maintenance. The nuclear plants of Ontario Hydro have evaluated on a number of alternative programs to improve their maintenance effectiveness and to reduce costs, including Reliability Centred Maintenance (RCM), call-up review, component-based PM programs, analysis of failure history and so on. Pickering ND (nuclear division) and Ontario Hydro's Nuclear Technologies Services Division, have embarked on a Condition Based Maintenance pilot project to address the above issues as a breakthrough solution for smarter maintenance. The Condition Based Maintenance pilot project will demonstrate an end-to-end process utilizing a Reliability Centred Maintenance structured approach to re-engineer and redefine the existing maintenance programs. The project emphasizes on-condition maintenance where justified, and utilizes an information management tool to provide the required records keeping and analysis infrastructure. This paper briefly describes the planned maintenance model at Pickering ND used to guide the CBM pilot, and an overview of the methodology used to develop on-condition equipment indicators as part of a re-engineered maintenance plan

Picker versus stripper harvesters on the High Plains of Texas

Science.gov (United States)

A break even analysis based on NPV was conducted to compare picker-based and stripper-based harvest systems with and without field cleaners. Under no conditions analyzed was the NPV of a stripper system without a field cleaner greater than a stripper system with a field cleaner. Break even curves re...

Pulse picker for synchrotron radiation driven by a surface acoustic wave.

Science.gov (United States)

Vadilonga, Simone; Zizak, Ivo; Roshchupkin, Dmitry; Petsiuk, Andrei; Dolbnya, Igor; Sawhney, Kawal; Erko, Alexei

2017-05-15

A functional test for a pulse picker for synchrotron radiation was performed at Diamond Light Source. The purpose of a pulse picker is to select which pulse from the synchrotron hybrid-mode bunch pattern reaches the experiment. In the present work, the Bragg reflection on a Si/B4C multilayer was modified using surface acoustic wave (SAW) trains. Diffraction on the SAW alters the direction of the x rays and it can be used to modulate the intensity of the x rays that reach the experimental chamber. Using electronic modulation of the SAW amplitude, it is possible to obtain different scattering conditions for different x-ray pulses. To isolate the single bunch, the state of the SAW must be changed in the short time gap between the pulses. To achieve the necessary time resolution, the measurements have been performed in conical diffraction geometry. The achieved time resolution was 120 ns.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1996

International Nuclear Information System (INIS)

1997-06-01

The Atomic Energy Control Board is the independent federal agency that controls all nuclear activities in Canada. A major use of nuclear energy in Canada is electricity production. The AECB assesses every station's performance against legal requirements, including the conditions in the operating licence. Each station is inspected and all aspects of the station's operation and management is reviewed. This report is the AECB staff assessment of reactor safety at the Pickering A and B Generating Stations for 1996. PNGS-A and PNGS-B operated safely during 1996. Although the risk to the workers and the public is low, major safety related changes are necessary at the stations and the sustainability of those changes needs to be demonstrated. Improvement is needed by Ontario Hydro in meeting the time limits for reporting reportable events. Ontario Hydro's follow up to events and causal factor analyses continue to need improvements. Improvements are needed to operational safety and reactor maintenance at both A and B. There are signs of improvement through Ontario Hydro's plan for recovery, and in station management changes. There also appears to be commitment to safety expressed at the highest level of the utility

Pipe support program at Pickering

International Nuclear Information System (INIS)

Sahazizian, L.A.; Jazic, Z.

1997-01-01

This paper describes the pipe support program at Pickering. The program addresses the highest priority in operating nuclear generating stations, safety. We present the need: safety, the process: managed and strategic, and the result: assurance of critical piping integrity. In the past, surveillance programs periodically inspected some systems, equipment, and individual components. This comprehensive program is based on a managed process that assesses risk to identify critical piping systems and supports and to develop a strategy for surveillance and maintenance. The strategy addresses all critical piping supports. Successful implementation of the program has provided assurance of critical piping and support integrity and has contributed to decreasing probability of pipe failure, reducing risk to worker and public safety, improving configuration management, and reducing probability of production losses. (author)

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Slack, J.; Norton, J.L.; Malkoske, G.R.

2003-01-01

therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

Automatic pickers performances in the case of the Emilia sequence of May-June 2012.

Science.gov (United States)

Tiberi, Lara; Spallarossa, Daniele; Costa, Giovanni

2013-04-01

The automatic processing of seismic data, whether for real-time seismic warning system or to reprocessing large amount of seismic recordings, is increasingly being demanded by seismologists especially in case of emergency as for the Emilia sequence in may-june 2012. In this study is presented a comparison between the AutoPicker (DipTeRiS, University of Genova) a new method used for automatic accurate onset phase picking for both P and S wave arrival based on the Akaike's information criterion (AIC), a solid and tested picker as the STA/LTA in Antelope software and the manual pickings. In order to construct the database used for the relocation of Emilia sequence, the RAN strong motion database has been merged with the available velocity and acceleration data extracted from the EIDA database (European Integrated Data Archive) and velocity data recorded by the Southeastern Alps Integrated Network (DMG, OGS, ARSO and ZAMG). The fault system of the Emilia earthquake area is complex and it is not easy to assess which fault has moved. A precise localization of the sequence is essential. The manual pickings, the equivalent locations and the choice of the most appropriate velocity model ("Iside") used in this study are the results of a work done in collaboration with Università di Chieti and DPC, not described here. The main problem of the AutoPicker and Antelope software is to discriminate events that occur very close to each other in time. The best way to solve that issue is choosing the best setup of both techniques to minimize the problem. Then we would like to implement the AutoPicker technique developed by Prof. Spallarossa on the Antelope system routinely used by UTS-DMG for the real-time data analysis.

A Study of Micro Finance: Special Reference to Female Waste Pickers in Pimpri Chinchwad Area in Pune

OpenAIRE

Hebalkar, Dr. Rashmi; Sharma, Meena Sunildutt

2013-01-01

Female waste pickers are the neglected section of urban women who are struggling to make ends meet, in an occupation which is hazardous for health, and are contributing to the welfare of society, without realizing it, through collecting waste and sending it forward for recycling. These women may be poorly educated but at least some of them have been unionized and their union attempts to improve their condition. Despite the existence of KKPKP union, there are female waste pickers who have not ...

Tritium releases from the Pickering Nuclear Generating Station and birth defects and infant mortality in nearby communities 1971-1988

International Nuclear Information System (INIS)

Johnson, K.C.; Rouleau, J.

1991-10-01

This study was commissioned to examine whether there were elevated rates of stillbirth, birth defects, or death in the first year of life between 1971 and 1988 among offspring of residents of communities within a 25-kilometre radius of the Pickering Nuclear Generating Station. The study was also to investigate whether there were any statistical associations between the monthly airborne or waterborne tritium emissions from the Pickering Nuclear Generating Station and the rates of these reproductive outcomes. Overall analysis did not support a hypothesis of increased rates of stillbirths, neonatal mortality or infant mortality near the Pickering Nuclear Generating Station, or a hypothesis of increased birth prevalence of birth defects for 21 of 22 diagnostic categories. The prevalence of Down Syndrome was elevated in both Pickering and Ajax; however, there was no consistent pattern between tritium release levels and Down Syndrome prevalence, chance could not be ruled out for the associations between Down Syndrome and tritium releases or ground-monitored concentrations, the association was detected in an analysis where multiple testing was done which may turn up significant associations by change, and maternal residence at birth and early in pregnancy needs to be verified. The association between Down Syndrome and low-level radiation remains indeterminate when existing evidence from epidemiological studies is summed. The estimated radiation exposure from the nuclear plant for residents of Pickering and Ajax is lower by a factor of 100 than the normal natural background radiation. Further study is recommended. (21 tabs., 29 figs., 5 maps, 37 refs.)

The interaction between popular economy, social movements and public policies: A case study of the waste pickers' movement

OpenAIRE

van Zeeland, Angelique J.W.M.

2014-01-01

This paper examines the challenges of expansion and sustainability of Social and Solidarity Economy (SSE). It focuses on the interaction between popular economy and SSE, and stresses the importance of collective action and public policies to enable the transition from the informal economy toward SSE. The main focus is on the waste pickers' movement. Experiences from Latin America, Asia and Africa show the possibilities of incorporating a significant contingent of informal waste pickers in sol...

Networks of recyclable material waste-picker's cooperatives: an alternative for the solid waste management in the city of Rio de Janeiro.

Science.gov (United States)

Tirado-Soto, Magda Martina; Zamberlan, Fabio Luiz

2013-04-01

The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city's main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers' cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro. Copyright © 2012 Elsevier Ltd. All rights reserved.

Preference of multi-walled carbon nanotube (MWCNT) to single-walled carbon nanotube (SWCNT) and activated carbon for preparing silica nanohybrid pickering emulsion for chemical enhanced oil recovery (C-EOR)

Energy Technology Data Exchange (ETDEWEB)

AfzaliTabar, M. [Department of Chemistry, Islamic Azad University Branch of Tehran North, Tehran (Iran, Islamic Republic of); Alaei, M., E-mail: alaiem@ripi.ir [Nanotechnology Research Center, Research Institute of Petroleum Industry (RIPI), Tehran (Iran, Islamic Republic of); Ranjineh Khojasteh, R.; Motiee, F. [Department of Chemistry, Islamic Azad University Branch of Tehran North, Tehran (Iran, Islamic Republic of); Rashidi, A.M. [Nanotechnology Research Center, Research Institute of Petroleum Industry (RIPI), Tehran (Iran, Islamic Republic of)

2017-01-15

The aim of this research was to determine the best nano hybrid that can be used as a Pickering emulsion Chemical Enhanced Oil Recovery (C-EOR). Therefore, we have prepared different carbon structures nano hybrids with SiO{sub 2} nano particles with different weight percent using sol-gel method. The as-prepared nano materials were characterized with X-Ray Diffraction (XRD), Field Emission Scanning Electron Microscopy (FE-SEM) and Thermal Gravimetric Analysis (TGA). Pickering emulsions of these nanohybrids were prepared at pH=7 in ambient temperature and with distilled water. Stability of the mentioned Pickering emulsions was controlled for one month. Emulsion phase morphology was investigated using optical microscopic imaging. Evaluation results demonstrated that the best sample is the 70% MWCNT/SiO{sub 2} nanohybrid. Stability of the selected nanohybrid (70% MWCNT/SiO{sub 2} nanohybrid) was investigated by alteration of salinity, pH and temperature. Results showed that the mentioned Pickering emulsion has very good stability at 0.1%, 1% salinity, moderate and high temperature (25 °C and 90 °C) and neutral and alkaline pH (7, 10) that is suitable for the oil reservoirs conditions. The effect of the related nano fluid on the wettability of carbonate rock was investigated by measuring the contact angle and interfacial tension. Results show that the nanofluid could significantly change the wettability of the carbonate rock from oil wet to water wet and can decrease the interfacial tension. Therefore, the 70% MWCNT/SiO{sub 2} nanohybrid Pickering emulsion can be used for Chemical Enhanced Oil Recovery (C-EOR).

Preference of multi-walled carbon nanotube (MWCNT) to single-walled carbon nanotube (SWCNT) and activated carbon for preparing silica nanohybrid pickering emulsion for chemical enhanced oil recovery (C-EOR)

International Nuclear Information System (INIS)

AfzaliTabar, M.; Alaei, M.; Ranjineh Khojasteh, R.; Motiee, F.; Rashidi, A.M.

2017-01-01

The aim of this research was to determine the best nano hybrid that can be used as a Pickering emulsion Chemical Enhanced Oil Recovery (C-EOR). Therefore, we have prepared different carbon structures nano hybrids with SiO 2 nano particles with different weight percent using sol-gel method. The as-prepared nano materials were characterized with X-Ray Diffraction (XRD), Field Emission Scanning Electron Microscopy (FE-SEM) and Thermal Gravimetric Analysis (TGA). Pickering emulsions of these nanohybrids were prepared at pH=7 in ambient temperature and with distilled water. Stability of the mentioned Pickering emulsions was controlled for one month. Emulsion phase morphology was investigated using optical microscopic imaging. Evaluation results demonstrated that the best sample is the 70% MWCNT/SiO 2 nanohybrid. Stability of the selected nanohybrid (70% MWCNT/SiO 2 nanohybrid) was investigated by alteration of salinity, pH and temperature. Results showed that the mentioned Pickering emulsion has very good stability at 0.1%, 1% salinity, moderate and high temperature (25 °C and 90 °C) and neutral and alkaline pH (7, 10) that is suitable for the oil reservoirs conditions. The effect of the related nano fluid on the wettability of carbonate rock was investigated by measuring the contact angle and interfacial tension. Results show that the nanofluid could significantly change the wettability of the carbonate rock from oil wet to water wet and can decrease the interfacial tension. Therefore, the 70% MWCNT/SiO 2 nanohybrid Pickering emulsion can be used for Chemical Enhanced Oil Recovery (C-EOR).

System dynamics applied to closed loop supply chains of desktops and laptops in Brazil: A perspective for social inclusion of waste pickers.

Science.gov (United States)

Ghisolfi, Verônica; Diniz Chaves, Gisele de Lorena; Ribeiro Siman, Renato; Xavier, Lúcia Helena

2017-02-01

The structure of reverse logistics for waste electrical and electronic equipment (WEEE) is essential to minimize the impacts of their improper disposal. In this context, the Brazilian Solid Waste Policy (BSWP) was a regulatory milestone in Brazil, submitting WEEE to the mandatory implementation of reverse logistics systems, involving the integration of waste pickers on the shared responsibility for the life cycle of products. This article aims to measure the impact of such legal incentives and the bargaining power obtained by the volume of collected waste on the effective formalization of waste pickers. The proposed model evaluates the sustainability of supply chains in terms of the use of raw materials due to disposal fees, collection, recycling and return of some materials from desktops and laptops using system dynamics methodology. The results show that even in the absence of bargaining power, the formalization of waste pickers occurs due to legal incentives. It is important to ensure the waste pickers cooperatives access to a minimum amount, which requires a level of protection against unfair competition with companies. Regarding the optimal level of environmental policies, even though the formalization time is long, it is still not enough to guarantee the formalization of waste picker cooperatives, which is dependent on their bargaining power. Steel is the material with the largest decrease in acquisition rate of raw material. Copyright © 2016 Elsevier Ltd. All rights reserved.

AECB staff annual report on Pickering NGS for the year 1989

International Nuclear Information System (INIS)

1990-06-01

This report presents a review of major licensing issues and the operational performance of Pickering NGS-A (Units 1-4) and Pickering NGS-B (Units 5-8) by the staff of the Atomic Energy Control Board (AECB) during 1989. Operations are monitored to ensure compliance with licensing requirements. This report is limited to those aspects that AECB staff consider to have particular safety significance. The number of infractions of the operating licence and the AECB Regulations doubled in 1989 compared to 1988. Three workers were exposed to radiation doses in excess of the regulatory limits. The AECB also found inadequate procedural compliance and an unacceptable level of housekeeping. Performance also requires improvement in response to AECB Health Physics appraisals; surveillance of systems by the Technical Section; chemical control performance; response to outstanding AECB action items; availability of special safety systems; operating memos, jumper records, call-ups and deficiency reports in effect; and fire fighting capability. Ontario Hydro has initiated a number of programs that are expected to result in improvements in these areas: an in-service station quality improvement plan; a program to improve and give assurance of compliance with the AECB Regulations, the operating licenses and the Operating Policies and Principles; a housekeeping and material condition improvement plan; and an action plan undertaken following radiation over-exposures. Given adequate resources and management support these programs should result in a noticeable improvement in station performance in 1990

Update of the Picker C9 irradiator control system of the gamma II room of the secondary laboratory of dosimetric calibration; Actualizacion del sistema de control del irradiador Picker C9 de la sala gamma II del laboratorio secundario de calibracion dosimetrica

Energy Technology Data Exchange (ETDEWEB)

Simon S, L. E.

2016-07-01

The Picker C9 irradiator is responsible for the calibration of different radiological equipment and the control system that maintains it in operation is designed in the graphical programming software LabVIEW (Laboratory Virtual Instrumentation Engineering Workbench), being its major advantages: the different types of communication, easy interconnection with other software and the recognition of different hardware devices, among others. Operation of the irradiator control system is performed with the NI-Usb-6008 (DAQ) data acquisition module of the National Instruments Company. The purpose of this work is to update the routines that make the Picker C9 control system of the gamma II room of the secondary laboratory of dosimetric calibration, using the graphic programming software LabVIEW, as well as to configure the new acquisition hardware of data that is implemented to control the Picker C9 irradiator system and ensure its operation. (Author)

Pickering Nuclear site wide groundwater monitoring system

International Nuclear Information System (INIS)

DeWilde, J.; Chin-Cheong, D.; Lledo, C.; Wootton, R.; Belanger, D.; Hansen, K.

2001-01-01

Ontario Power Generation Inc. (OPG) is continuing its efforts to understand the chemical and physical characteristics of the groundwater flow systems beneath the Pickering Nuclear Generating Station (PNGS). To this end, OPG constructed a site-wide Groundwater Monitoring System (GMS) at the PNGS to provide support to other ongoing environmental investigations and to provide a means to monitor current and future groundwater environmental issues. This paper will present the results of this work, including the development of a state-of-the-art data management system for storage and retrieval of environmental data for the site, which has applications for other power generation facilities. (author)

Viscosity of the oil-in-water Pickering emulsion stabilized by surfactant-polymer and nanoparticle-surfactant-polymer system

Science.gov (United States)

Sharma, Tushar; Kumar, G. Suresh; Chon, Bo Hyun; Sangwai, Jitendra S.

2014-11-01

Information on the viscosity of Pickering emulsion is required for their successful application in upstream oil and gas industry to understand their stability at extreme environment. In this work, a novel formulation of oil-in-water (o/w) Pickering emulsion stabilized using nanoparticle-surfactant-polymer (polyacrylamide) system as formulated in our earlier work (Sharma et al., Journal of Industrial and Engineering Chemistry, 2014) is investigated for rheological stability at high pressure and high temperature (HPHT) conditions using a controlled-strain rheometer. The nanoparticle (SiO2 and clay) concentration is varied from 1.0 to 5.0 wt%. The results are compared with the rheological behavior of simple o/w emulsion stabilized by surfactant-polymer system. Both the emulsions exhibit non-Newtonian shear thinning behavior. A positive shift in this behavior is observed for surfactant-polymer stabilized emulsion at high pressure conditions. Yield stress is observed to increase with pressure for surfactant-polymer emulsion. In addition, increase in temperature has an adverse effect on the viscosity of emulsion stabilized by surfactant-polymer system. In case of nanoparticle-surfactant-polymer stabilized o/w emulsion system, the viscosity and yield stress are predominantly constant for varying pressure and temperature conditions. The viscosity data for both o/w emulsion systems are fitted by the Herschel-Bulkley model and found to be satisfactory. In general, the study indicates that the Pickering emulsion stabilized by nanoparticle-surfactant-polymer system shows improved and stable rheological properties as compared to conventional emulsion stabilized by surfactant-polymer system indicating their successful application for HPHT environment in upstream oil and gas industry.

Pickering emulsions stabilized by whey protein nanoparticles prepared by thermal cross-linking

NARCIS (Netherlands)

Wu, Jiande; Shi, Mengxuan; Li, Wei; Zhao, Luhai; Wang, Ze; Yan, Xinzhong; Norde, Willem; Li, Yuan

2015-01-01

A Pickering (o/w) emulsion was formed and stabilized by whey protein isolate nanoparticles (WPI NPs). Those WPI NPs were prepared by thermal cross-linking of denatured WPI proteins within w/o emulsion droplets at 80. °C for 15. min. During heating of w/o emulsions containing 10% (w/v) WPI

Fabrication of CMC-g-PAM Superporous Polymer Monoliths via Eco-Friendly Pickering-MIPEs for Superior Adsorption of Methyl Violet and Methylene Blue.

Science.gov (United States)

Wang, Feng; Zhu, Yongfeng; Wang, Wenbo; Zong, Li; Lu, Taotao; Wang, Aiqin

2017-01-01

A series of superporous carboxymethylcellulose- graft -poly(acrylamide)/palygorskite (CMC- g -PAM/Pal) polymer monoliths presenting interconnected pore structure and excellent adsorption properties were prepared by one-step free-radical grafting polymerization reaction of CMC and acrylamide (AM) in the oil-in-water (O/W) Pickering-medium internal phase emulsions (Pickering-MIPEs) composed of non-toxic edible oil as a dispersion phase and natural Pal nanorods as stabilizers. The effects of Pal dosage, AM dosage, and co-surfactant Tween-20 (T-20) on the pore structures of the monoliths were studied. It was revealed that the well-defined pores were formed when the dosages of Pal and T-20 are 9-14 and 3%, respectively. The porous monolith can rapidly adsorb 1,585 mg/g of methyl violet (MV) and 1,625 mg/g of methylene blue (MB). After the monolith was regenerated by adsorption-desorption process for five times, the adsorption capacities still reached 92.1% (for MV) and 93.5% (for MB) of the initial maximum adsorption capacities. The adsorption process was fitted with Langmuir adsorption isotherm model and pseudo-second-order adsorption kinetic model very well, which indicate that mono-layer chemical adsorption mainly contribute to the high-capacity adsorption for dyes. The superporous polymer monolith prepared from eco-friendly Pickering-MIPEs shows good adsorption capacity and fast adsorption rate, which is potential adsorbent for the decontamination of dye-containing wastewater.

Fabrication of CMC-g-PAM Superporous Polymer Monoliths via Eco-Friendly Pickering-MIPEs for Superior Adsorption of Methyl Violet and Methylene Blue

Directory of Open Access Journals (Sweden)

Feng Wang

2017-06-01

Full Text Available A series of superporous carboxymethylcellulose-graft-poly(acrylamide/palygorskite (CMC-g-PAM/Pal polymer monoliths presenting interconnected pore structure and excellent adsorption properties were prepared by one-step free-radical grafting polymerization reaction of CMC and acrylamide (AM in the oil-in-water (O/W Pickering-medium internal phase emulsions (Pickering-MIPEs composed of non-toxic edible oil as a dispersion phase and natural Pal nanorods as stabilizers. The effects of Pal dosage, AM dosage, and co-surfactant Tween-20 (T-20 on the pore structures of the monoliths were studied. It was revealed that the well-defined pores were formed when the dosages of Pal and T-20 are 9–14 and 3%, respectively. The porous monolith can rapidly adsorb 1,585 mg/g of methyl violet (MV and 1,625 mg/g of methylene blue (MB. After the monolith was regenerated by adsorption-desorption process for five times, the adsorption capacities still reached 92.1% (for MV and 93.5% (for MB of the initial maximum adsorption capacities. The adsorption process was fitted with Langmuir adsorption isotherm model and pseudo-second-order adsorption kinetic model very well, which indicate that mono-layer chemical adsorption mainly contribute to the high-capacity adsorption for dyes. The superporous polymer monolith prepared from eco-friendly Pickering-MIPEs shows good adsorption capacity and fast adsorption rate, which is potential adsorbent for the decontamination of dye-containing wastewater.

High-Surface-Area, Emulsion-Templated Carbon Foams by Activation of polyHIPEs Derived from Pickering Emulsions

Directory of Open Access Journals (Sweden)

Robert T. Woodward

2016-09-01

Full Text Available Carbon foams displaying hierarchical porosity and excellent surface areas of >1400 m2/g can be produced by the activation of macroporous poly(divinylbenzene. Poly(divinylbenzene was synthesized from the polymerization of the continuous, but minority, phase of a simple high internal phase Pickering emulsion. By the addition of KOH, chemical activation of the materials is induced during carbonization, producing Pickering-emulsion-templated carbon foams, or carboHIPEs, with tailorable macropore diameters and surface areas almost triple that of those previously reported. The retention of the customizable, macroporous open-cell structure of the poly(divinylbenzene precursor and the production of a large degree of microporosity during activation leads to tailorable carboHIPEs with excellent surface areas.

Waste picker livelihoods and inclusive neoliberal municipal solid waste management policies: The case of the La Chureca garbage dump site in Managua, Nicaragua.

Science.gov (United States)

Hartmann, Chris

2018-01-01

The modernization (i.e. mechanization, formalization, and capital intensification) and enclosure of municipal solid waste management (MSWM) systems threaten waste picker livelihoods. From 2009 to 2013, a major development project, embodying traditional neoliberal policies with inclusive social policies, transformed the Managua, Nicaragua, municipal solid waste site from an open-air dump where as many as 2,000 informal waste pickers toiled to a sanitary landfill. To investigate waste pickers' social and economic condition, including labor characteristics, household income, and poverty incidence, after the project's completion, 146 semi-structured survey questionnaires were administered to four communities adjacent to the landfill and 45 semi-structured interviews were completed with key stakeholders. Findings indicate that hundreds of waste pickers were displaced by the project, employment benefits from the project were unevenly distributed by neighborhood, and informal waste picking endures due to persistent impoverishment, thereby contributing to continued social and economic marginalization and environmental degradation. The findings highlight the limitations of inclusive neoliberal development efforts to transform MSWM in a low-income country. Copyright © 2017 Elsevier Ltd. All rights reserved.

Expanding worldwide urban solid waste recycling: The Brazilian social technology in waste pickers inclusion.

Science.gov (United States)

Rutkowski, Jacqueline E; Rutkowski, Emília W

2015-12-01

'If an integrated urban waste management system includes the informal recycling sector (IRS), there is a good chance that more solid waste is recycled' is common sense. However, informal integration brings additional social, environmental, and economic benefits, such as reduction of operational costs and environmental impacts of landfilling. Brazil is a global best practice example in terms of waste picker inclusion, and has received international recognition for its recycling levels. In addition to analysing the results of inclusive recycling approaches, this article evaluates a selection of the best Brazilian inclusive recycling practices and summaries and presents the resulting knowledge. The objective is to identify processes that enable the replication of the inclusion of the informal recycling sector model as part of municipal solid waste management. Qualitative and quantitative data have been collected in 25 Brazilian cities that have contracted waste pickers co-operatives for door-to-door selective collection of recyclables. Field data was collected in action research projects that worked with waste pickers co-operatives between 2006 and 2013. The Brazilian informal recycling sector integration model improves municipal solid waste recycling indicators: it shows an increase in the net tonness recycled, from 140 to 208 t month(-1), at a much lower cost per tonne than conventional selective collection systems. Inclusive systems show costs of US$35 per tonne of recyclables collected, well below the national average of US$195.26. This inclusive model improves the quality of collected material and the efficiency of municipal selective collection. It also diminishes the negative impacts of informal recycling, by reducing child labour, and by improving the conditions of work, occupational health and safety, and uncontrolled pollution. Although treating the Brazilian experience as a blueprint for transfer of experience in every case is unrealistic, the results

Pickering emulsion stabilized by cashew gum- poly-l-lactide copolymer nanoparticles: Synthesis, characterization and amphotericin B encapsulation.

Science.gov (United States)

Richter, A R; Feitosa, J P A; Paula, H C B; Goycoolea, F M; de Paula, R C M

2018-04-01

In this work, we provide proof-of-concept of formation, physical characteristics and potential use as a drug delivery formulation of Pickering emulsions (PE) obtained by a novel method that combines nanoprecipitation with subsequent spontaneous emulsification process. To this end, pre-formed ultra-small (d.∼10 nm) nanoprecipitated nanoparticles of hydrophobic derivatives of cashew tree gum grafted with polylactide (CGPLAP), were conceived to stabilize Pickering emulsions obtained by spontaneous emulsification. These were also loaded with Amphotericin B (AmB), a drug of low oral bioavailability used in the therapy of neglected diseases such as leishmaniasis. The graft reaction was performed in two CG/PLA molar ratio conditions (1:1 and 1:10). Emulsions were prepared by adding the organic phase (Miglyol 812 ® ) in the aqueous phase (nanoprecipitated CGPLAP), resulting the immediate emulsion formation. The isolation by centrifugation does not destabilize or separate the nanoparticles from oil droplets of the PE emulsion. Emulsions with CGPLAP 1:1 presented unimodal distributions at different CGPLA concentration, lower values in size and PDI and the best stability over time. The AmB was incorporated in the emulsions with a process efficiency of 21-47%, as determined by UV-vis. AmB in CGPLAP emulsions is in less aggregated state than observed in commercial AmB formulation. Copyright © 2018 Elsevier B.V. All rights reserved.

Multi-Unit Aspects of the Pickering Generating Station

Energy Technology Data Exchange (ETDEWEB)

Morison, W. G. [Atomic Energy of Canada Ltd, Sheridan Park, ON (Canada)

1968-04-15

The Pickering nuclear generating station is located on the north shore of Lake Ontario, about 20 miles east of the city of Toronto, Canada. The station has been planned and laid out on an eight-unit station, four units of which have now been authorized for construction. Each of these four units consists of a single heavy-water moderated and cooled CANDU-type reactor and auxiliaries coupled to a single tandem compound turbine generator with a net output of approximately 500 MW(e). The units are identical and are scheduled to come into operation at intervals of one year from 1970 to 1973. The station has been planned with central facilities for: administration maintenance laboratories, stores, change rooms, decontamination and waste management services. A common control centre, cooling water intake and discharge system, and spent fuel storage bay for four units has been arranged. A feature of the multi-unit station is a common containment system. Cost savings in building a number of identical units on the same site result from a single exclusion area, shared engineering costs, equipment purchase contracts for four identical components, and efficient use of construction plant. Operating cost savings are anticipated in the use of a common operating and maintenance staff and spare parts inventory. The plant has been arranged to minimize problems of operating, commissioning and constructing units at the same time on the same site. The layout and construction sequence have been arranged so that the first unit can be commissioned and operated with little or no interference from the construction forces working on succeeding units. During the construction phase barriers will be erected in the common control centre between operating control equipment and that being installed. Operations and construction personnel will enter the plant by separate routes and work in areas separated by physical barriers. (author)

The attached algae community near Pickering GS: III

International Nuclear Information System (INIS)

McKinley, S.R.

1982-01-01

The relationship between attached algae and macro-invertebrates in the nearshore zone of Lake Ontario was investigated in the vicinity of the Pickering 'A' NGS. Measures of faunal density, richness, evenness, and biomass were generally higher from areas which supported attached algae. Gammarus fasciatus, Cricotopus bicinctus, Dicrotendipes spp., Orthocladius obumbratus, Cladotanytarsus spp., Orthocladius spp., and Parakiefferiella spp., were significantly correlated with algal standing crop. All of the above dominant invertebrates ingested epiphytes associated with Cladophora glomerata. Attempts to explain the distribution of the zoobenthic assemblages using the physical/biological characteristics of the study area indicated algal cover, substrate size, wind velocity and water temperature were most important

Update of the Picker C9 irradiator control system of the gamma II room of the secondary laboratory of dosimetric calibration

International Nuclear Information System (INIS)

Simon S, L. E.

2016-01-01

The Picker C9 irradiator is responsible for the calibration of different radiological equipment and the control system that maintains it in operation is designed in the graphical programming software LabVIEW (Laboratory Virtual Instrumentation Engineering Workbench), being its major advantages: the different types of communication, easy interconnection with other software and the recognition of different hardware devices, among others. Operation of the irradiator control system is performed with the NI-Usb-6008 (DAQ) data acquisition module of the National Instruments Company. The purpose of this work is to update the routines that make the Picker C9 control system of the gamma II room of the secondary laboratory of dosimetric calibration, using the graphic programming software LabVIEW, as well as to configure the new acquisition hardware of data that is implemented to control the Picker C9 irradiator system and ensure its operation. (Author)

Seismic response of the Pickering pressure relief duct to the 1985 Nahanni earthquake

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objective of this study is to examine the structural response of the Pickering pressure relief duct when subjected to the ground motion records of the 1985 Nahanni earthquake (December 23, 05:16 GMT, Site 1 - Iverson, N.W.T.). It also includes an estimate of the possible impact on the nuclear safety function of the duct. The structural models developed in an earlier study were used in this analysis. The response to the earthquake ground motion was determined on the basis of the estimated capacities of various components of the duct. The ability of the structure to fulfill its nuclear safety function is discussed. (author). 6 refs., 1 tab., 17 figs

Interfacial behaviour of sodium stearoyllactylate (SSL) as an oil-in-water pickering emulsion stabiliser.

Science.gov (United States)

Kurukji, D; Pichot, R; Spyropoulos, F; Norton, I T

2013-11-01

The ability of a food ingredient, sodium stearoyllactylate (SSL), to stabilise oil-in-water (O/W) emulsions against coalescence was investigated, and closely linked to its capacity to act as a Pickering stabiliser. Results showed that emulsion stability could be achieved with a relatively low SSL concentration (≥0.1 wt%), and cryogenic-scanning electron microscopy (cryo-SEM) visualisation of emulsion structure revealed the presence of colloidal SSL aggregates adsorbed at the oil-water interface. Surface properties of SSL could be modified by altering the size of these aggregates in water; a faster decrease in surface tension was observed when SSL dispersions were subjected to high pressure homogenisation (HPH). The rate of SSL adsorption at the sunflower oil-water interface also increased after HPH, and a higher interfacial tension (IFT) was observed with increasing SSL concentration. Differential scanning calorimetry (DSC) enabled a comparison of the thermal behaviour of SSL in aqueous dispersions with SSL-stabilised O/W emulsions. SSL melting enthalpy depended on emulsion interfacial area and the corresponding DSC data was used to determine the amount of SSL adsorbed at the oil-water interface. An idealised theoretical interfacial coverage calculation based on Pickering emulsion theory was in general agreement with the mass of SSL adsorbed as predicted by DSC. Copyright © 2013 The Authors. Published by Elsevier Inc. All rights reserved.

Effects of Occupational Exposure on the Health of Rag Pickers Due to Fungal Contamination at Waste Dumping Sites in Gwalipor (India

Directory of Open Access Journals (Sweden)

Harandra K. Sharma

2017-02-01

Full Text Available We investigated fungal contamination near different waste dumping sites and assessed the health risk factors of rag pickers associated with collection of waste in Gwalior during the year 2014-15. Petri plates were exposed at waste dumping sites and were transferred to the laboratory, analysis and identification was mainly carried out by culturing the fungal colonies by following standard procedures. A pretested questionnaire was used to evaluate the health problems among the rag pickers. Results indicated that all the dumping sites are contaminated with different types of fungal pathogens like Alternaria alternate, Aspergillus flavus, A. fumigates, A. niger, Cladosporium, Fusarium, Mucor, Penicillium and Rhizopus. Our study reported higher incidence of musculoskeletal and respiratory diseases among rag pickers. There is also strong need for carrying out similar assessment studies for other cities too. This will entail generation of more precise site specific information regarding fungal species and associated health risk factor.

Effects of Occupational Exposure on the Health of Rag Pickers Due to Fungal Contamination at Waste Dumping Sites in Gwalior (India

Directory of Open Access Journals (Sweden)

Harandra K. Sharma

2017-02-01

Full Text Available We investigated fungal contamination near different waste dumping sites and assessed the health risk factors of rag pickers associated with collection of waste in Gwalior during the year 2014-15. Petri plates were exposed at waste dumping sites and were transferred to the laboratory, analysis and identification was mainly carried out by culturing the fungal colonies by following standard procedures. A pretested questionnaire was used to evaluate the health problems among the rag pickers. Results indicated that all the dumping sites are contaminated with different types of fungal pathogens like Alternaria alternate, Aspergillus flavus, A. fumigates, A. niger, Cladosporium, Fusarium, Mucor, Penicillium and Rhizopus. Our study reported higher incidence of musculoskeletal and respiratory diseases among rag pickers. There is also strong need for carrying out similar assessment studies for other cities too. This will entail generation of more precise site specific information regarding fungal species and associated health risk factor.

Hydroxyapatite-armored poly(ε-caprolactone) microspheres and hydroxyapatite microcapsules fabricated via a Pickering emulsion route.

Science.gov (United States)

Fujii, Syuji; Okada, Masahiro; Nishimura, Taiki; Maeda, Hayata; Sugimoto, Tatsuya; Hamasaki, Hiroyuki; Furuzono, Tsutomu; Nakamura, Yoshinobu

2012-05-15

Hydroxyapatite (HAp) nanoparticle-armored poly(ε-caprolactone) (PCL) microspheres were fabricated via a "Pickering-type" emulsion solvent evaporation method in the absence of any molecular surfactants. It was clarified that the interaction between carbonyl/carboxylic acid groups of PCL and the HAp nanoparticles at an oil-water interface played a crucial role in the preparation of the stable Pickering-type emulsions and the HAp nanoparticle-armored microspheres. The HAp nanoparticle-armored PCL microspheres were characterized in terms of size, size distribution, morphology, and chemical compositions using scanning electron microscopy, laser diffraction, energy dispersive X-ray microanalysis, and thermogravimetric analysis. The presence of HAp nanoparticles at the surface of the microspheres was confirmed by scanning electron microscopy and energy dispersive X-ray microanalysis. Pyrolysis of the PCL cores led to the formation of the corresponding HAp hollow microcapsules. Copyright © 2012 Elsevier Inc. All rights reserved.

Quebec Gentilly 2 nuclear power station

International Nuclear Information System (INIS)

Labbe, J.A.

Modifications and commissioning of the Gentilly reactor are described. The Gentilly reactor is owned by AECL, not Quebec Hydro, and has served as a prototype reactor. The Gentilly-2 reactor is a 'packaged' 600 MWe PHW reactor similar to Pickering-1, etc. Interesting aspects of construction and purchasing of equipment are described. (E.C.B.)

Fabrication of CMC-g-PAM superporous polymer monoliths via eco-friendly Pickering-MIPEs for superior adsorption of methyl violet and methylene blue

Science.gov (United States)

Wang, Feng; Zhu, Yongfeng; Wang, Wenbo; Zong, Li; Lu, Taotao; Wang, Aiqin

2017-06-01

A series of superporous carboxymethylcellulose-graft-poly(acrylamide) (CMC-g-PAM) polymer monoliths presenting interconnected pore structure and excellent adsorption properties were prepared by one-step free-radical grafting polymerization reaction of CMC and acrylamide (AM) in the oil-in-water (O/W) Pickering-medium internal phase emulsions (Pickering-MIPEs) composed of non-toxic edible oil as a dispersion phase and natural Pal nanorods as stabilizers. The effects of Pal dosage, AM dosage, and co-surfactant Tween-20 (T-20) on the pore structures of the monoliths were studied. It was revealed that the well-defined pores were formed when the dosages of Pal and T-20 are 9-14% and 3%, respectively. The porous monolith can rapidly adsorb 1585 mg/g of methyl violet (MV) and 1625 mg/g of methylene blue (MB). After the monolith was regenerated by adsorption-desorption process for 5 times, the adsorption capacities still reached 92.1% (for MV) and 93.5% (for MB) of the initial maximum adsorption capacities. The adsorption process was fitted with Langmuir adsorption isotherm model and pseudo-second-order adsorption kinetic model very well, which indicate that mono-layer chemical adsorption mainly contribute to the high-capacity adsorption for dyes. The superporous polymer monolith prepared from eco-friendly Pickering-MIPEs shows good adsorption capacity and fast adsorption rate, which is potential adsorbent for the decontimination of dye-containing wastewater.

Pickering irradiated fuel transfer conveyor isolation

Energy Technology Data Exchange (ETDEWEB)

Koivisto, D J; Eijsermans, L J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1997-12-31

Pickering A NGS has been in operation for 25 years and is one of the longest in service CANDU stations. Some underwater fuel handling equipment, notably the conveyor stops, have been without maintenance throughout that time. This paper describes the concept of a conveyor isolation system that permits draining of a single or multiple elevator columns and also the early stages of a development program for the elastomeric sealing element. The prototype seal element has been proven in lab tests to be capable of limiting leakage to 0.5 IGPM (imperial gallons per minute) at the design pressure of 6.5 psi. The design of a sealing element is particularly interesting because the conveyor tube is a square cross-section which contains an additional obstruction , a conveyor drive cable. A seal delivery, actuating and positioning system has been conceptually laid out and the design is proceeding, with projected implementation in 1998. (author). 8 figs.

Pickering irradiated fuel transfer conveyor isolation

International Nuclear Information System (INIS)

Koivisto, D.J.; Eijsermans, L.J.

1996-01-01

Pickering A NGS has been in operation for 25 years and is one of the longest in service CANDU stations. Some underwater fuel handling equipment, notably the conveyor stops, have been without maintenance throughout that time. This paper describes the concept of a conveyor isolation system that permits draining of a single or multiple elevator columns and also the early stages of a development program for the elastomeric sealing element. The prototype seal element has been proven in lab tests to be capable of limiting leakage to 0.5 IGPM (imperial gallons per minute) at the design pressure of 6.5 psi. The design of a sealing element is particularly interesting because the conveyor tube is a square cross-section which contains an additional obstruction , a conveyor drive cable. A seal delivery, actuating and positioning system has been conceptually laid out and the design is proceeding, with projected implementation in 1998. (author). 8 figs

Uplifting : mobile cranes and picker trucks get higher, stronger, and safer

Energy Technology Data Exchange (ETDEWEB)

Budd, G.

2008-10-15

New crane and picker truck equipment designed for use in the oil and gas industry was discussed in this article. The longest crane in the North America is due to arrive in Calgary soon. Mobile cranes are often used for maintenance, construction, and shutdowns at natural gas plants, refineries, and oil batteries. Telescopic cranes are also used to service pump jacks and lift rolls of coiled tubing into place on drilling rigs. While cranes carry more weight, picker trucks are more mobile and flexible. Lattice boom crawler cranes were designed to pick up loads and carry them to different positions. Telescopic cranes must remain stationary, and careful planning is needed to ensure that they are used efficiently. Modern telescopic cranes have hydraulically powered booms. Advanced steel alloy technology is used to produce lightweight booms equipped with automated pin-locking mechanisms. The largest telescopic crane in North America will be used by an oil and gas operator in Alberta. The crane will have a combined boom and lattice jib height of 226 meters, and its maximum lifting capacity is 1200 tonnes. Gantry cranes are also increasingly being used by oil and gas operators due to their ability to straddle loads, thereby reducing the risk of tipping. It was concluded that gantry cranes are particularly suitable for rougher terrains. 4 figs.

Eddy current magnetic bias x-probe qualification and inspection of steam generator Monel 400 tubing in Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Lepine, B.A.; Van Langen, J.; Obrutsky, L.

2006-01-01

This paper presents an overview of the x-probe MB 350 eddy current inspection array probe, for detection of open OD axial crack-like flaws in Monel 400 tubes at Pickering Nuclear Generating Station. This report contains a selection of inspection results from the field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, specially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential out diameter (OD) axially oriented crack. Post-inspection tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability μ r Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

Online control loop tuning in Pickering Nuclear Generating Stations

International Nuclear Information System (INIS)

Yu, K.X.; Harrington, S.

2008-01-01

Most analog controllers in the Pickering B Nuclear Generating Stations adopted PID control scheme. In replacing the analog controllers with digital controllers, the PID control strategies, including the original tuning parameters were retained. The replacement strategy resulted in minimum effort on control loop tuning. In a few cases, however, it was found during commissioning that control loop tuning was required as a result of poor control loop performance, typically due to slow response and controlled process oscillation. Several factors are accounted for the necessities of control loop re-tuning. Our experience in commissioning the digital controllers showed that online control tuning posted some challenges in nuclear power plant. (author)

Arresting relaxation in Pickering Emulsions

Science.gov (United States)

Atherton, Tim; Burke, Chris

2015-03-01

Pickering emulsions consist of droplets of one fluid dispersed in a host fluid and stabilized by colloidal particles absorbed at the fluid-fluid interface. Everyday materials such as crude oil and food products like salad dressing are examples of these materials. Particles can stabilize non spherical droplet shapes in these emulsions through the following sequence: first, an isolated droplet is deformed, e.g. by an electric field, increasing the surface area above the equilibrium value; additional particles are then adsorbed to the interface reducing the surface tension. The droplet is then allowed to relax toward a sphere. If more particles were adsorbed than can be accommodated by the surface area of the spherical ground state, relaxation of the droplet is arrested at some non-spherical shape. Because the energetic cost of removing adsorbed colloids exceeds the interfacial driving force, these configurations can remain stable over long timescales. In this presentation, we present a computational study of the ordering present in anisotropic droplets produced through the mechanism of arrested relaxation and discuss the interplay between the geometry of the droplet, the dynamical process that produced it, and the structure of the defects observed.

Histidine-functionalized carbon-based dot-Zinc(II) nanoparticles as a novel stabilizer for Pickering emulsion synthesis of polystyrene microspheres.

Science.gov (United States)

Ruiyi, Li; Zaijun, Li; Junkang, Liu

2017-05-01

Carbon-based dots (CDs) are nanoparticles with size-dependent optical and electronic properties that have been widely applied in energy-efficient displays and lighting, photovoltaic devices and biological markers. However, conventional CDs are difficult to be used as ideal stabilizer for Pickering emulsion due to its irrational amphiphilic structure. The study designed and synthesized a new histidine-functionalized carbon dot-Zinc(II) nanoparticles, which is termed as His-CD-Zn. The His-CD was made via one-step hydrothermal treatment of histidine and maleic acid. The His-CD reacted with Zn 2+ to form His-CD-Zn. The as-prepared His-CD-Zn was used as a solid particle surfactant for stabilizing styrene-in-water emulsion. The Pickering emulsion exhibits high stability and sensitive pH-switching behaviour. The introduction of S 2 O 8 2- triggers the emulsion polymerization of styrene. The resulted polystyrene microsphere was well coated with His-CDs on the surface. It was successfully used as an ideal adsorbent for removal of heavy metallic ions from water with high adsorption capacity. The study also provides a prominent approach for fabrication of amphiphilic carbon-based nanoparticles for stabilizing Pickering emulsion. Copyright © 2017 Elsevier Inc. All rights reserved.

Pickering G.S. boiler repair: an example of planned maintenance

International Nuclear Information System (INIS)

Dalrymple, D.G.

1976-04-01

The first application of boiler repair tools and procedures is estimated to have yielded a four-fold return on the development investment. The need to develop such technology is a result of the environment in which boiler repairs must be made. As nuclear technology evolves and plants and components get bigger, equipment will increasingly have to be repaired in situ with minimum plant downtime and minimum exposure of repair personnel to radiation. This lecture traces development of the Pickering A boiler repair capability which is seen as an example of how utility and contractor should interact to anticipate and meet maintenance requirements. (author)

Pickering NGS emergency water supply system emergency start flow simulation and experiment

Energy Technology Data Exchange (ETDEWEB)

Davidge, E.; Misra, A. [Ontario Power Generation Inc., Nuclear Safety Analysis & Technology Department, Toronto, Ontario (Canada)

2012-07-01

A proposed modification to the OPG Pickering Nuclear Generation Station Emergency Water Supply (EWS) system was analyzed using the Industry Standard Toolset code GOTHIC to determine the acceptability of the proposed system configuration during pump start-up. The new configuration of the system included a vertical dead-ended pipe, initially filled with air. The simulation demonstrated that no significant water hammer effects were predicted and tests performed with the new configuration confirmed the analysis results. (author)

Phytotoxicology section investigation in the vicinity of the Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the Darlington Nuclear Generating Station, in October, 1989

International Nuclear Information System (INIS)

1991-02-01

The Phytotoxicology Section, Air Resources Branch is a participant in the Pickering and Bruce Nuclear Contingency Plans. The Phytotoxicology Emergency Response Team is responsible for collecting vegetation samples in the event of a nuclear emergency at any of the nuclear generating stations in the province. As part of its responsibility the Phytotoxicology Section collects samples around the nuclear generating stations for comparison purposes in the event of an emergency. Because of the limited frequency of sampling, the data from the surveys are not intended to be used as part of a regulatory monitoring program. These data represent an effort by the MOE to begin to establish a data base of tritium concentrations in vegetation. The Phytotoxicology Section has carried out seven surveys in the vicinity of Ontario Hydro nuclear generating stations since 1981. Surveys were conducted for tritium in snow in the vicinity of Bruce Nuclear Power Development (BNPD), February, 1981; tritium in cell-free water of white ash in the vicinity of BNPD, September, 1981; tritium in snow in the vicinity of BNPD, March, 1982; tritium in tree sap in the vicinity of BNPD, April, 1982; tritium in tree sap in the vicinity of BNPD, April, 1984, tritium in the cell-free water of white ash in the vicinity of BNPD, September, 1985; and, tritium in cell-free water of grass in the vicinity of Pickering Nuclear Generation Station (PNGS), October 1986. In all cases a pattern of decreasing tritium levels with increasing distance from the stations was observed. In October, 1989, assessment surveys were conducted around Bruce Nuclear Power Development, the Pickering Nuclear Generating Station and the new Darlington Nuclear Generating Station (DNGS). The purpose of these surveys was to provide baseline data for tritium in cell-free water of grass at all three locations at the same time of year. As none of the reactor units at DNGS had been brought on line at the time of the survey, this data was to be

Waste Picker Organizations and Their Contribution to the Circular Economy: Two Case Studies from a Global South Perspective

Directory of Open Access Journals (Sweden)

Jutta Gutberlet

2017-09-01

Full Text Available The discussion on the circular economy (CE has attracted a rising interest within global policy and business as a way of increasing the sustainability of production and consumption. Yet the literature mostly portrays a Global North perspective. There is a diverse spectrum of community-based organizations playing important roles in resource recovery and transformation, particularly, but not only, in Global South countries, providing innovative examples for grassroots involvement in waste management and in the CE. This article proposes to add a Southern lens, situated in the context of waste picker organizations, to the concept of CE. The discursive framework in this article couples ecological economy (EE with social/solidarity economy (SSE, focusing not only on environmental sustainability but also on social, economic, political and cultural dimensions involved in production, consumption and discard. We acknowledge that grassroots movements contribute to policy making and improve urban waste management systems. The paper outlines two empirical studies (Argentina, Brazil that illustrate how waste picker organizations perform selective waste collection services, engage with municipalities and industries, and practice the CE. The research reveals that social and political facets need to be added to the debate about the CE, linking environmental management and policy with community development and recognizing waste pickers as protagonists in the CE. Our findings emphasize a need for a change of persisting inequalities in public policy by recognizing the importance of popular waste management praxis and knowledge, ultimately redefining the CE.

Colloidal formulations for probiotics delivery and Pickering systems

DEFF Research Database (Denmark)

Yücel Falco, Cigdem

countries. One emerging functional food area is the efficient delivery of health-promoting probiotics. Although much progress has already been made in the development and understanding of novel microencapsulation systems, maintaining viability during gastric passage and being effective at the target site...... is still an issue for probiotics. On the other hand, one of the foremost challenges in the production of physically stable foods during the defined shelf life is the identification of new food-grade ingredients. In this context, the replacement of classical emulsifiers with solid particles is one...... of the advancing food research areas, though the number of food-grade solid particles investigated is still insufficient. Edible probiotic strains can potentially be valorised as particles similar to micron-sized fat particles in Pickering systems such as ice cream due to their low calories and their availability...

Reverse logistics network for municipal solid waste management: The inclusion of waste pickers as a Brazilian legal requirement

International Nuclear Information System (INIS)

Ferri, Giovane Lopes; Diniz Chaves, Gisele de Lorena; Ribeiro, Glaydston Mattos

2015-01-01

Highlights: • We propose a reverse logistics network for MSW involving waste pickers. • A generic facility location mathematical model was validated in a Brazilian city. • The results enable to predict the capacity for screening and storage centres (SSC). • We minimise the costs for transporting MSW with screening and storage centres. • The use of SSC can be a potential source of revenue and a better use of MSW. - Abstract: This study proposes a reverse logistics network involved in the management of municipal solid waste (MSW) to solve the challenge of economically managing these wastes considering the recent legal requirements of the Brazilian Waste Management Policy. The feasibility of the allocation of MSW material recovery facilities (MRF) as intermediate points between the generators of these wastes and the options for reuse and disposal was evaluated, as well as the participation of associations and cooperatives of waste pickers. This network was mathematically modelled and validated through a scenario analysis of the municipality of São Mateus, which makes the location model more complete and applicable in practice. The mathematical model allows the determination of the number of facilities required for the reverse logistics network, their location, capacities, and product flows between these facilities. The fixed costs of installation and operation of the proposed MRF were balanced with the reduction of transport costs, allowing the inclusion of waste pickers to the reverse logistics network. The main contribution of this study lies in the proposition of a reverse logistics network for MSW simultaneously involving legal, environmental, economic and social criteria, which is a very complex goal. This study can guide practices in other countries that have realities similar to those in Brazil of accelerated urbanisation without adequate planning for solid waste management, added to the strong presence of waste pickers that, through the

Reverse logistics network for municipal solid waste management: The inclusion of waste pickers as a Brazilian legal requirement

Energy Technology Data Exchange (ETDEWEB)

Ferri, Giovane Lopes, E-mail: giovane.ferri@aluno.ufes.br [Department of Engineering and Technology, Federal University of Espírito Santo – UFES, Rodovia BR 101 Norte, Km 60, Bairro Litorâneo, São Mateus, ES, 29.932-540 (Brazil); Diniz Chaves, Gisele de Lorena, E-mail: gisele.chaves@ufes.br [Department of Engineering and Technology, Federal University of Espírito Santo – UFES, Rodovia BR 101 Norte, Km 60, Bairro Litorâneo, São Mateus, ES, 29.932-540 (Brazil); Ribeiro, Glaydston Mattos, E-mail: glaydston@pet.coppe.ufrj.br [Transportation Engineering Programme, Federal University of Rio de Janeiro – UFRJ, Centro de Tecnologia, Bloco H, Sala 106, Cidade Universitária, Rio de Janeiro, 21949-900 (Brazil)

2015-06-15

Highlights: • We propose a reverse logistics network for MSW involving waste pickers. • A generic facility location mathematical model was validated in a Brazilian city. • The results enable to predict the capacity for screening and storage centres (SSC). • We minimise the costs for transporting MSW with screening and storage centres. • The use of SSC can be a potential source of revenue and a better use of MSW. - Abstract: This study proposes a reverse logistics network involved in the management of municipal solid waste (MSW) to solve the challenge of economically managing these wastes considering the recent legal requirements of the Brazilian Waste Management Policy. The feasibility of the allocation of MSW material recovery facilities (MRF) as intermediate points between the generators of these wastes and the options for reuse and disposal was evaluated, as well as the participation of associations and cooperatives of waste pickers. This network was mathematically modelled and validated through a scenario analysis of the municipality of São Mateus, which makes the location model more complete and applicable in practice. The mathematical model allows the determination of the number of facilities required for the reverse logistics network, their location, capacities, and product flows between these facilities. The fixed costs of installation and operation of the proposed MRF were balanced with the reduction of transport costs, allowing the inclusion of waste pickers to the reverse logistics network. The main contribution of this study lies in the proposition of a reverse logistics network for MSW simultaneously involving legal, environmental, economic and social criteria, which is a very complex goal. This study can guide practices in other countries that have realities similar to those in Brazil of accelerated urbanisation without adequate planning for solid waste management, added to the strong presence of waste pickers that, through the

Pickering emulsions stabilized by biodegradable block copolymer micelles for controlled topical drug delivery.

Science.gov (United States)

Laredj-Bourezg, Faiza; Bolzinger, Marie-Alexandrine; Pelletier, Jocelyne; Chevalier, Yves

2017-10-05

Surfactant-free biocompatible and biodegradable Pickering emulsions were investigated as vehicles for skin delivery of hydrophobic drugs. O/w emulsions of medium-chain triglyceride (MCT) oil droplets loaded with all-trans retinol as a model hydrophobic drug were stabilized by block copolymer nanoparticles: either poly(lactide)-block-poly(ethylene glycol) (PLA-b-PEG) or poly(caprolactone)-block-poly(ethylene glycol) (PCL-b-PEG). Those innovative emulsions were prepared using two different processes allowing drug loading either inside oil droplets or inside both oil droplets and non-adsorbed block copolymer nanoparticles. Skin absorption of retinol was investigated in vitro on pig skin biopsies using the Franz cell method. Supplementary experiments by confocal fluorescence microscopy allowed the visualization of skin absorption of the Nile Red dye on histological sections. Retinol and Nile Red absorption experiments showed the large accumulation of hydrophobic drugs in the stratum corneum for the Pickering emulsions compared to the surfactant-based emulsion and an oil solution. Loading drug inside both oil droplets and block copolymer nanoparticles enhanced again skin absorption of drugs, which was ascribed to the supplementary contribution of free block copolymer nanoparticles loaded with drug. Such effect allowed tuning drug delivery to skin over a wide range by means of a suitable selection of either the formulation or the drug loading process. Copyright © 2017 Elsevier B.V. All rights reserved.

Selective removal of erythromycin by magnetic imprinted polymers synthesized from chitosan-stabilized Pickering emulsion.

Science.gov (United States)

Ou, Hongxiang; Chen, Qunhui; Pan, Jianming; Zhang, Yunlei; Huang, Yong; Qi, Xueyong

2015-05-30

Magnetic imprinted polymers (MIPs) were synthesized by Pickering emulsion polymerization and used to adsorb erythromycin (ERY) from aqueous solution. The oil-in-water Pickering emulsion was stabilized by chitosan nanoparticles with hydrophobic Fe3O4 nanoparticles as magnetic carrier. The imprinting system was fabricated by radical polymerization with functional and crosslinked monomer in the oil phase. Batches of static and dynamic adsorption experiments were conducted to analyze the adsorption performance on ERY. Isotherm data of MIPs well fitted the Freundlich model (from 15 °C to 35 °C), which indicated heterogeneous adsorption for ERY. The ERY adsorption capacity of MIPs was about 52.32 μmol/g at 15 °C. The adsorption kinetics was well described by the pseudo-first-order model, which suggested that physical interactions were primarily responsible for ERY adsorption. The Thomas model used in the fixed-bed adsorption design provided a better fit to the experimental data. Meanwhile, ERY exhibited higher affinity during adsorption on the MIPs compared with the adsorption capacity of azithromycin and chloramphenicol. The MIPs also exhibited excellent regeneration capacity with only about 5.04% adsorption efficiency loss in at least three repeated adsorption-desorption cycles. Copyright © 2015 Elsevier B.V. All rights reserved.

Investigation of tritium in groundwater at Pickering NGS

International Nuclear Information System (INIS)

DeWilde, J.; Yu, L.; Belanger, D.; Wootton, R.; Hansen, K.; McGurk, E.; Teare, A.

2001-01-01

Ontario Power Generation Inc. (OPG) investigated tritium in groundwater at the Pickering Nuclear Generating Station (PNGS). The objectives of the study were to evaluate and define the extent of radio-nuclides, primarily tritium, in groundwater, investigate the causes or sources of contamination, determine impacts on the natural environment, and provide recommendations to prevent future discharges. This paper provides an overview of the investigations conducted in 1999 and 2000 to identify the extent of the tritium beneath the site and the potential sources of tritium released to the groundwater. The investigation and findings are summarized with a focus on unique aspects of the investigation, on lessons learned and benefits. Some of the investigative techniques discussed include process assessments, video inspections, hydrostatic and tracer tests, Helium 3 analysis for tritium age dating, deuterium and tritium in soil analysis. The investigative techniques have widespread applications to other nuclear generating stations. (author)

Training courses at VR-1 reactor

International Nuclear Information System (INIS)

Sklenka, L.; Kropik, M.

2006-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

Current status of the Thai Research Reactor (TRR-1/M1)

International Nuclear Information System (INIS)

Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

2006-01-01

The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

The Vale of Pickering in the Mesolithic: uncovering the early post-glacial landscape

Directory of Open Access Journals (Sweden)

Tim Schadla-Hall

2000-10-01

Full Text Available Since 1954, when Grahame Clark published the results of his excavations at Star Carr in northeast Yorkshire, the site has been recognized as a key to understanding early Mesolithic huntergatherer settlement and subsistence in northwest Europe. In 1976, archaeological and palaeoenvironmental research in the area was resumed - since 1986 under the auspices of the Vale of Pickering Research Trust - and it is now possible to set Star Carr and nearby Mesolithic sites in the wider context of the early postglacial landscape.

Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

Energy Technology Data Exchange (ETDEWEB)

Montes, J

1972-07-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Assessment of ichthyoplankton entrainment at Pickering 'A' NGS using a pump/net in lake system

International Nuclear Information System (INIS)

McKinley, R.S.

1985-03-01

Annual entrainment at Pickering 'A' NGS was estimated for alewife as 13.6 X 10 6 larvae and 409 X 10 6 eggs. A substantial portion of eggs and larvae entering the intake were dead due to natural mortality (41%-81%) prior to entrainment. Viable eggs and larvae, immediately following entrainment showed mortalities of 54% and 44% respectively. The latent mortality of entrained eggs was 100% (48 h)

Remote tooling for inspection and repair in Pickering NGS-A calandria vault

International Nuclear Information System (INIS)

Hadji-Mirzai, M.; Tokarz, A.; Vandenberg, J.P.

1993-01-01

In recent years it has been necessary to develop capabilities for the inspection and repair of carbon steel components located within calandria vaults at Ontario Hydro's Pickering Nuclear Generating Station 'A'. Concerns about corrosion of piping and some of the structural components have made necessary the development of remote manipulators to inspect and repair carbon steel components within the vaults to ensure continued reliable operation of the units. Remote manipulators for this program have been designed to perform a number of inspection and repair tasks, and several versions have been developed to specialise in detailed inspection techniques and precision tooling module manipulation. (author)

NUKAB system use with the PICKER DYNA CAMERA II

International Nuclear Information System (INIS)

Collet, H.; Faurous, P.; Lehn, A.; Suquet, P.

Present-day data processing units connected to scintillation gamma cameras can make use of cabled programme or recorded programme systems. The NUKAB system calls on the latter technique. The central element of the data processing unit, connected to the PICKER DYNA CAMERA II output, consists of a DIGITAL PDP 8E computer with 12-bit technological words. The use of a 12-bit technological format restricts the possibilities of digitalisation, 64x64 images representing the practical limit. However the NUKAB system appears well suited to the processing of data from gamma cameras at present in service. The addition of output terminals of the tracing panel type should widen the possibilities of the system. It seems that the 64x64 format is not a handicap in view of the resolution power of the detectors [fr

Development and applications of reactor noise analysis at Ontario Hydro's CANDU reactors

International Nuclear Information System (INIS)

Gloeckler, O.; Tulett, M.V.

1995-01-01

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro's CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Nozzleless Fabrication of Oil-Core Biopolymeric Microcapsules by the Interfacial Gelation of Pickering Emulsion Templates.

Science.gov (United States)

Leong, Jun-Yee; Tey, Beng-Ti; Tan, Chin-Ping; Chan, Eng-Seng

2015-08-05

Ionotropic gelation has been an attractive method for the fabrication of biopolymeric oil-core microcapsules due to its safe and mild processing conditions. However, the mandatory use of a nozzle system to form the microcapsules restricts the process scalability and the production of small microcapsules (microcapsules through ionotropic gelation at the interface of an O/W Pickering emulsion. This approach involves the self-assembly of calcium carbonate (CaCO3) nanoparticles at the interface of O/W emulsion droplets followed by the addition of a polyanionic biopolymer into the aqueous phase. Subsequently, CaCO3 nanoparticles are dissolved by pH reduction, thus liberating Ca(2+) ions to cross-link the surrounding polyanionic biopolymer to form a shell that encapsulates the oil droplet. We demonstrate the versatility of this method by fabricating microcapsules from different types of polyanionic biopolymers (i.e., alginate, pectin, and gellan gum) and water-immiscible liquid cores (i.e., palm olein, cyclohexane, dichloromethane, and toluene). In addition, small microcapsules with a mean size smaller than 100 μm can be produced by selecting the appropriate conventional emulsification methods available to prepare the Pickering emulsion. The simplicity and versatility of this method allows biopolymeric microcapsules to be fabricated with ease by ionotropic gelation for numerous applications.

Annual report on JEN-1 reactor

International Nuclear Information System (INIS)

Montes, J.

1972-01-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Development and applications of reactor noise analysis at Ontario Hydro`s CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O [Ontario Hydro, Toronto, ON (Canada); Tulett, M V [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro`s CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Reverse logistics network for municipal solid waste management: The inclusion of waste pickers as a Brazilian legal requirement.

Science.gov (United States)

Ferri, Giovane Lopes; Chaves, Gisele de Lorena Diniz; Ribeiro, Glaydston Mattos

2015-06-01

This study proposes a reverse logistics network involved in the management of municipal solid waste (MSW) to solve the challenge of economically managing these wastes considering the recent legal requirements of the Brazilian Waste Management Policy. The feasibility of the allocation of MSW material recovery facilities (MRF) as intermediate points between the generators of these wastes and the options for reuse and disposal was evaluated, as well as the participation of associations and cooperatives of waste pickers. This network was mathematically modelled and validated through a scenario analysis of the municipality of São Mateus, which makes the location model more complete and applicable in practice. The mathematical model allows the determination of the number of facilities required for the reverse logistics network, their location, capacities, and product flows between these facilities. The fixed costs of installation and operation of the proposed MRF were balanced with the reduction of transport costs, allowing the inclusion of waste pickers to the reverse logistics network. The main contribution of this study lies in the proposition of a reverse logistics network for MSW simultaneously involving legal, environmental, economic and social criteria, which is a very complex goal. This study can guide practices in other countries that have realities similar to those in Brazil of accelerated urbanisation without adequate planning for solid waste management, added to the strong presence of waste pickers that, through the characteristic of social vulnerability, must be included in the system. In addition to the theoretical contribution to the reverse logistics network problem, this study aids in decision-making for public managers who have limited technical and administrative capacities for the management of solid wastes. Copyright © 2015 Elsevier Ltd. All rights reserved.

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

2003-01-01

Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

Electrospun composite matrices of poly(ε-caprolactone)-montmorillonite made using tenside free Pickering emulsions

International Nuclear Information System (INIS)

Samanta, Archana; Takkar, Sonam; Kulshreshtha, Ritu; Nandan, Bhanu; Srivastava, Rajiv K.

2016-01-01

The production of composite electrospun matrices of poly(ε-caprolactone) (PCL) using an emulsifier-free emulsion, made with minimal organic solvent, as precursor is reported. Pickering emulsions of PCL were prepared using modified montmorillonite (MMT) clay as the stabilizer. Hydrophobic tallow group of the modified MMT clay resulted in analogous interaction of clay with oil and aqueous phase and its adsorption at the interface to provide stability to the resultant emulsion. Composite fibrous matrices of PCL and MMT were produced using electrospinning under controlled conditions. The fiber fineness was found to alter with PCL concentration and volume fraction of the aqueous and oil phases. A higher tensile strength and modulus was obtained with inclusion of MMT in PCL electrospun matrix in comparison to a matrix made using neat PCL. The presence of clay in the fibrous matrix did not change the cell proliferation efficiency in comparison to neat PCL matrix. Composite fibrous matrices of PCL/MMT bearing enhanced tensile properties may find applications in areas other than tissue engineering for example food packaging and filtration. - Highlights: • Tenside free, clay stabilized Pickering emulsion of PCL is made with minimal organic solvent. • Organic–inorganic composite fibrous matrices were produced via emulsion electrospinning. • Fiber fineness was efficiently controlled by variation in emulsion formulation. • Fibrous matrices of high tensile strength and modulus were obtained in comparison to neat PCL matrix. • PCL/clay matrices showed effective cell proliferation as a neat PCL matrix.

Electrospun composite matrices of poly(ε-caprolactone)-montmorillonite made using tenside free Pickering emulsions

Energy Technology Data Exchange (ETDEWEB)

Samanta, Archana [Department of Textile Technology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India); Takkar, Sonam; Kulshreshtha, Ritu [Department of Biochemical Engineering and Biotechnology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India); Nandan, Bhanu [Department of Textile Technology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India); Srivastava, Rajiv K., E-mail: rajiv@textile.iitd.ac.in [Department of Textile Technology, Indian Institute of Technology Delhi, Hauz Khas, New Delhi 110016 (India)

2016-12-01

The production of composite electrospun matrices of poly(ε-caprolactone) (PCL) using an emulsifier-free emulsion, made with minimal organic solvent, as precursor is reported. Pickering emulsions of PCL were prepared using modified montmorillonite (MMT) clay as the stabilizer. Hydrophobic tallow group of the modified MMT clay resulted in analogous interaction of clay with oil and aqueous phase and its adsorption at the interface to provide stability to the resultant emulsion. Composite fibrous matrices of PCL and MMT were produced using electrospinning under controlled conditions. The fiber fineness was found to alter with PCL concentration and volume fraction of the aqueous and oil phases. A higher tensile strength and modulus was obtained with inclusion of MMT in PCL electrospun matrix in comparison to a matrix made using neat PCL. The presence of clay in the fibrous matrix did not change the cell proliferation efficiency in comparison to neat PCL matrix. Composite fibrous matrices of PCL/MMT bearing enhanced tensile properties may find applications in areas other than tissue engineering for example food packaging and filtration. - Highlights: • Tenside free, clay stabilized Pickering emulsion of PCL is made with minimal organic solvent. • Organic–inorganic composite fibrous matrices were produced via emulsion electrospinning. • Fiber fineness was efficiently controlled by variation in emulsion formulation. • Fibrous matrices of high tensile strength and modulus were obtained in comparison to neat PCL matrix. • PCL/clay matrices showed effective cell proliferation as a neat PCL matrix.

Fragger: a protein fragment picker for structural queries.

Science.gov (United States)

Berenger, Francois; Simoncini, David; Voet, Arnout; Shrestha, Rojan; Zhang, Kam Y J

2017-01-01

Protein modeling and design activities often require querying the Protein Data Bank (PDB) with a structural fragment, possibly containing gaps. For some applications, it is preferable to work on a specific subset of the PDB or with unpublished structures. These requirements, along with specific user needs, motivated the creation of a new software to manage and query 3D protein fragments. Fragger is a protein fragment picker that allows protein fragment databases to be created and queried. All fragment lengths are supported and any set of PDB files can be used to create a database. Fragger can efficiently search a fragment database with a query fragment and a distance threshold. Matching fragments are ranked by distance to the query. The query fragment can have structural gaps and the allowed amino acid sequences matching a query can be constrained via a regular expression of one-letter amino acid codes. Fragger also incorporates a tool to compute the backbone RMSD of one versus many fragments in high throughput. Fragger should be useful for protein design, loop grafting and related structural bioinformatics tasks.

CANDU fuel - fifteen years of power reactor experience

International Nuclear Information System (INIS)

Fanjoy, G.R.; Bain, A.S.

1977-01-01

CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

Exploring the role of picker personality in predicting picking performance with pick by voice, pick to light and RF-terminal picking

NARCIS (Netherlands)

de Vries, J.; de Koster, R.; Stam, D.

2016-01-01

Order pickers and individual differences between them could have a substantial impact on picking performance, but are largely ignored in studies on order picking. This paper explores the role of individual differences in picking performance with various picking tools (pick by voice, RF-terminal

Preparation of stable Pickering emulsions with short, medium and long chain fats and starch nanocrystals and their in vitro digestion properties

Science.gov (United States)

Pickering emulsions are receiving more attention as delivery systems in food and pharmaceuticals because they can be formulated with nontoxic food ingredients to form stable emulsions. In this study, 40-100 nm starch nanocrystals (SNCs) prepared from acid hydrolysis of waxy maize starches were used ...

Maintenance of ageing CANDU reactors. A regulatory perspective

International Nuclear Information System (INIS)

Dunstan, T.

1996-01-01

The subject of this paper is, 'requirements for maintenance of ageing reactors from the perspective of a regulator', with a focus on the particular theme of; 'continuing safety assurance'. A major role of maintenance is to ensure the continuing reliability and effectiveness of safety related systems and equipment. Continuing safety assurance is an issue the Atomic Energy Control Board has been wrestling with for some time. From my perspective, much remains to be done before the AECB can be confident that Canadian nuclear plants have the necessary programs in place to achieve continuing safety assurance. To introduce the topic, it would be appropriate to say a few words about the AECB's position with respect to the situation at the Pickering NGS. Why did we blow the whistle last August and, what are we doing about it? (author)

Moving ring reactor 'Karin-1'

International Nuclear Information System (INIS)

1983-12-01

The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

Energy Technology Data Exchange (ETDEWEB)

Barnett, P. C. [Canadian General Electric (Canada)

1968-07-15

The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

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

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

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

Occupational analysis for the Angra-1 reactor

International Nuclear Information System (INIS)

Moraes, A.

1991-01-01

Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

Reactor operations at SAFARI-1

International Nuclear Information System (INIS)

Vlok, J.W.H.

2003-01-01

A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

Tritium in groundwater investigation at the Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

DeWilde, J.; Yu, L.; Wootton, R.; Belanger, D.; Hansen, K.; McGurk, E.; Teare, A.

2001-01-01

Ontario Power Generation Inc. (OPG) investigated tritium in groundwater at the Pickering Nuclear Generating Station (PNGS). The objectives of the study were to evaluate and define the extent of radionuclides, primarily tritium, in groundwater, investigate the causes or sources of contamination, determine impacts on the natural environment, and provide recommendations to prevent future discharges. This paper provides an overview of the investigations conducted in 1999 and 2000 to identity the extent of the tritium beneath the site and the potential sources of tritium released to the groundwater. The investigation and findings are summarized with a focus on unique aspects of the investigation, on lessons learned and benefits. Some of the investigative techniques discussed include process assessments, video inspections, hydrostatic and tracer tests, Helium 3 analysis for tritium age dating, deuterium and tritium in soil analysis. The investigative techniques have widespread applications to other nuclear generating stations. (author)

Molecularly imprinted polymer microspheres prepared by Pickering emulsion polymerization for selective solid-phase extraction of eight bisphenols from human urine samples

International Nuclear Information System (INIS)

Yang, Jiajia; Li, Yun; Wang, Jincheng; Sun, Xiaoli; Cao, Rong; Sun, Hao; Huang, Chaonan; Chen, Jiping

2015-01-01

Highlights: • BPA imprinted polymer microspheres were prepared by Pickering emulsion polymerization. • Regular spherical shape and narrow diameter distribution. • Good specific adsorption capacity for BPA. • Good class-selectivity and clean-up efficiency for bisphenols in human urine under SPE mode. • Good recoveries and sensitivity for bisphenols using the MIPMS-SPE coupled with HPLC-DAD method. - Abstract: The bisphenol A (BPA) imprinted polymer microspheres were prepared by simple Pickering emulsion polymerization. Compared to traditional bulk polymerization, both high yields of polymer and good control of particle sizes were achieved. The characterization results of scanning electron microscopy and nitrogen adsorption–desorption measurements showed that the obtained molecularly imprinted polymer microsphere (MIPMS) particles possessed regular spherical shape, narrow diameter distribution (30–60 μm), a specific surface area (S BET ) of 281.26 m 2 g −1 and a total pore volume (V t ) of 0.459 cm 3 g −1 . Good specific adsorption capacity for BPA was obtained in the sorption experiment and good class selectivity for BPA and its seven structural analogs (bisphenol F, bisphenol B, bisphenol E, bisphenol AF, bisphenol S, bisphenol AP and bisphenol Z) was demonstrated by the chromatographic evaluation experiment. The MIPMS as solid-phase extraction (SPE) packing material was then evaluated for extraction and clean-up of these bisphenols (BPs) from human urine samples. An accurate and sensitive analytical method based on the MIPMS-SPE coupled with HPLC-DAD has been successfully established for simultaneous determination of eight BPs from human urine samples with detection limits of 1.2–2.2 ng mL −1 . The recoveries of BPs for urine samples at two spiking levels (100 and 500 ng mL −1 for each BP) were in the range of 81.3–106.7% with RSD values below 8.3%

Glacial Lake Pickering: stratigraphy and chronology of a proglacial lake dammed by the North Sea Lobe of the British-Irish Ice Sheet

OpenAIRE

Evans, David J.A.; Bateman, Mark D.; Roberts, David H.; Medialdea, Alicia; Hayes, Laura; Duller, Geoff A.T.; Fabel, Derek; Clark, Chris D.

2016-01-01

We report the first chronology, using four new optically stimulated luminescence dates, on the sedimentary record of Glacial Lake Pickering, dammed by the North Sea Lobe of the British–Irish Ice Sheet during the Dimlington Stadial (24–11 ka cal BP). Dates range from 17.6 ± 1.0 to 15.8 ± 0.9 ka for the sedimentation of the Sherburn Sands at East Heslerton, which were formed by multiple coalescing alluvial fans prograding into the falling water levels of the lake and fed by progressively larger...

A progress review of Ontario Hydro's nuclear generation and heavy water production programs

International Nuclear Information System (INIS)

Kee, F.J.; Woodhead, L.W.

Performance and economics of CANDU reactors in service are described. Progress of commissioning, construction and planning of reactors at Pickering, Bruce, and Darlington is outlined. Heavy water production is reviewed. (E.C.B.)

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Karel, Matejka; Lubomir, Sklenka

2005-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

Fuel channel life limiting factors that dictate fuel channel maintenance requirements

International Nuclear Information System (INIS)

Richinson, P.J.; Wong, H.W.; Ellis, P.J.

1995-01-01

CANDU reactors have been operating for 33 years. The Nuclear Power Demonstration (NPD) Unit started up in 1962 and the prototype of CANDU, Douglas Point, started in 1967. The first commercial reactors, Pickering Units 1 and 2 both went into service in 1971 closely followed by Units 3 and 4 in 1972 and 1973 respectively. Operating commercial reactor experience represents over 10,000 pressure tubes, not including the replaced channels in all the Pickering A Units, and nearly 130,000 pressure tube operating years. No pressure tube has yet operated for its 30 year design lifetime of 210 KEFPH at 80% capacity factor. The longest operating time for pressure tubes to-date is about 120 KEFPH in Pickering Unit 4. Many lessons have been learned regarding pressure tube life limiting factors from the early CANDU units and these, together with the information obtained from an extensive pressure tube R and D program, have resulted in many design changes and improvements in material properties, mainly from manufacturing route changes. Reactors built recently are expected to achieve their 30 year design life. The development of Periodic and In-service Inspection programs and equipment, assessment methodologies and acceptance criteria, and the development of maintenance tooling and procedures are enabling the life limiting factors to be addressed in the currently operating units. The life limiting factors in currently operating Units are reviewed in relation to the experience gained from the early units, the R and D programs and the inspection and maintenance performed to date. (author)

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

The Effect of Storage and Routing Policies on Picker Blocking in a Real-life Narrow-aisle Warehouse

OpenAIRE

Van Gils, Teun; Caris, An; Ramaekers, Katrien

2017-01-01

Upcoming e-commerce markets force warehouses to handle a large number of orders within short time windows. Narrow-aisle order picking systems allow to store a large number of products in small areas. In manual order picking systems, narrow aisles can result in substantial waiting time compared to wide-aisle systems. The objective of this study is to analyse the joint effect of the two main operational order picking planning problems, storage location assignment and order picker routing, on or...

Characteristics of outage radiation fields around various reactor components

International Nuclear Information System (INIS)

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

2008-01-01

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

The case for new nuclear

International Nuclear Information System (INIS)

Luxat, J.C.

2013-01-01

Over a 22 year period from 1971 to 1993 a total of 20 reactor units were brought into service - an average of approximately one unit per year. Ontario Hydro constructed the four-unit Pickering A station, four units at Bruce A, four units at Pickering B, four units at Bruce B and four units at Darlington during this period. This represents a capacity of nearly 14,000 MW, as shown in Figure 1. During this period there was a large increase in industrial capacity in Ontario, particularly in manufacturing, driven in large measure by the incentives offered by low electricity prices, skilled workers and a good health care system. Subsequently in the mid-1990's the Pickering A and Bruce A units were laid up and maintenance efforts were focused on the Pickering B, Bruce B and Darlington stations. Two of the four units at Pickering A were returned to service in the early 2000's and the four units of Bruce A were returned to service with two units being refurbished. By 2010 nuclear capacity in the province had returned to 12,800 MW. The Ontario Long Term Energy Plan (LTEP) announced at the beginning of December does not include new build nuclear but does include refurbishment of the Darlington station as well as two units at Bruce A and four units at Bruce B. The six units at Pickering will be shut down by 2020. As shown in Figure 1, this will reduce the nuclear capacity from the current 12,800 MW to 8000 MW when the Pickering A and B units are removed from service in 2020 and the refurbishment of Darlington and Bruce units proceeds starting in 2016 and projected to complete by 2031. This will be the lowest nuclear generating capacity in the province since 1985. (author)

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

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

1994-01-01

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

Burnup measurements at the RECH-1 research reactor

International Nuclear Information System (INIS)

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

2002-01-01

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

Interconnectivity of macroporous molecularly imprinted polymers fabricated by hydroxyapatite-stabilized Pickering high internal phase emulsions-hydrogels for the selective recognition of protein.

Science.gov (United States)

Sun, Yanhua; Li, Yuqing; Xu, Jiangfeng; Huang, Ling; Qiu, Tianyun; Zhong, Shian

2017-07-01

Hydroxyapatite hybridized molecularly imprinted polydopamine polymers with selective recognition of bovine hemoglobin (BHb) were successfully prepared via Pickering oil-in-water high internal phase emulsions-hydrogels and molecularly imprinting technique. The emulsions were stabilized by hydroxyapatite of which the wettability was modified by 3-methacryloxypropyltrimethoxysilane. The materials were characterized by SEM, IR and TGA. The results showed that the BHb imprinted polymers based on Pickering hydrogels (Hydro-MIPs) possess macropores ranging from 20μm to 50μm, and their large numbers of amino groups and hydroxyl groups result in a favorable adsorption capacity for BHb. The maximum adsorption capacity of Hydro-MIPs for BHb was 438mg/g, 3.27 times more than that of the non-imprinted polymers (Hydro-NIPs). The results indicated that Hydro-MIPs possessing well-defined hierarchical porous structures exhibited outstanding recognition behavior towards the target protein molecules. This work provided a promising alternative method for the fabrication of polymer materials with tunable and interconnected pores structures for the separation and purification of protein in vitro. Copyright © 2017. Published by Elsevier B.V.

Intestinal parasitism among waste pickers in Mato Grosso do Sul, Midwest Brazil

Directory of Open Access Journals (Sweden)

Minoru German Higa Júnior

2017-12-01

Full Text Available ABSTRACT The purpose of this study was to estimate the prevalence of intestinal parasites in both cooperative-affiliated and independent waste pickers operating at the municipal sanitary landfill in Campo Grande, Mato Grosso do Sul, Brazil, and associate these findings with hemoglobin, eosinophils, vitamin A and C levels and interleukin 5 and 10 (IL-5 and IL-10 production. Biological samples were collected, in addition to clinical, epidemiological, and sociodemographic data. Stool analyzes were based on sedimentation by centrifugation and on spontaneous sedimentation. High-performance liquid chromatography was used to determine vitamin A and C levels. ELISA was employed to quantify interleukins. Intestinal parasites were found in 29 of the 66 subjects assessed (43.9%. Endolimax nana (22.7%, Entamoeba coli (21.1%, Giardia lamblia (6.1%, Entamoeba histolytica/E. dispar (4.5%, and Ascaris lumbricoides (4.5% were the most prevalent species. Pathogenic parasites were detected in 11 individuals (16.7%. Hypovitaminoses A and C were detected in 19.6% (13/66 and 98.4% (65/66 of subjects, respectively. IL-5 and IL-10 production was observed in 21 (31.8% and 32 (48.4% subjects, respectively. Infection with pathogenic intestinal parasites was not a cause of vitamin A and C deficiency or IL-5 and IL-10 production among these workers.

A comparison of clinical vs subclinical skin pickers in Israel.

Science.gov (United States)

Keuthen, Nancy J; Curley, Erin E; Tung, Esther S; Ittah, Karen; Qasem, Atheer; Murad, Sari; Odlaug, Brian L; Leibovici, Vera

2016-05-01

Skin-picking disorder (SPD) was recognized as its own entity for the first time in DSM-5. The existing SPD literature is limited and, to date, no study has examined the differences between clinical and sub- clinical SPD. Identifying differences between these 2 groups may improve diagnostic accuracy, treatment, and prevention efforts. Israeli adults (N = 4,325) from 2 previous studies were examined for the presence of clinical and subclinical SPD. Individuals with clinical SPD (n = 150) vs subclinical SPD (n = 219) were compared on skin-picking characteristics, psychological phenomena, and clinical correlates. There were many similarities between clinical and subclinical skin pickers. Individuals with clinical SPD, however, had more severe skin picking, greater associated functional impairment, greater perceived stress, and greater depressive and obsessive-compulsive symptoms, and were also more likely to have a first-degree relative with SPD. This study suggests that although there are some similarities between clinical and subclinical SPD, there also are distinct differences in the clinical presentation. Understanding these differences may be an important factor in treatment and prevention planning.

Laboratory testing and assessment of the Pickering PRD supporting frame

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objective of this study was to design and test reinforced concrete beam-column subassemblages representing the beam, column and joint of the Centre Pier (CP) support of the Pressure Relief Duct (PRD) at the Pickering A Nuclear Generating Station. The testing program was expected to establish the failure mode of the subassemblage and to compare the performance of the existing CP with a specimen detailed in accordance with current code provisions. A one-third scale specimen of the beam-column subassemblage was designed and tested to failure when subjected to simulated seismic loads. A second specimen was constructed with shear reinforcement that was detailed according to the provisions of the CAN3-N287.3-M82 code. The second specimen was tested in the same manner as the first specimen. From the experimental data on the behaviour and mode of failure of the specimens, analytical evaluations were conducted to determine the inelastic nonlinear behaviour of the CP structural system when subjected to various levels of ground motion. (author). 11 refs., 3 tabs., 40 figs

Molecularly imprinted polymer microspheres prepared by Pickering emulsion polymerization for selective solid-phase extraction of eight bisphenols from human urine samples

Energy Technology Data Exchange (ETDEWEB)

Yang, Jiajia [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Li, Yun; Wang, Jincheng [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); Sun, Xiaoli; Cao, Rong [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Sun, Hao [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); Department of Chemistry, Liaoning University, Shenyang 110000 (China); Huang, Chaonan [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Chen, Jiping, E-mail: chenjp@dicp.ac.cn [Key Laboratory of Separation Sciences for Analytical Chemistry, Dalian Institute of Chemical Physics, Chinese Academy of Sciences, 457 Zhongshan Road, Dalian 116023 (China)

2015-05-04

Highlights: • BPA imprinted polymer microspheres were prepared by Pickering emulsion polymerization. • Regular spherical shape and narrow diameter distribution. • Good specific adsorption capacity for BPA. • Good class-selectivity and clean-up efficiency for bisphenols in human urine under SPE mode. • Good recoveries and sensitivity for bisphenols using the MIPMS-SPE coupled with HPLC-DAD method. - Abstract: The bisphenol A (BPA) imprinted polymer microspheres were prepared by simple Pickering emulsion polymerization. Compared to traditional bulk polymerization, both high yields of polymer and good control of particle sizes were achieved. The characterization results of scanning electron microscopy and nitrogen adsorption–desorption measurements showed that the obtained molecularly imprinted polymer microsphere (MIPMS) particles possessed regular spherical shape, narrow diameter distribution (30–60 μm), a specific surface area (S{sub BET}) of 281.26 m{sup 2} g{sup −1} and a total pore volume (V{sub t}) of 0.459 cm{sup 3} g{sup −1}. Good specific adsorption capacity for BPA was obtained in the sorption experiment and good class selectivity for BPA and its seven structural analogs (bisphenol F, bisphenol B, bisphenol E, bisphenol AF, bisphenol S, bisphenol AP and bisphenol Z) was demonstrated by the chromatographic evaluation experiment. The MIPMS as solid-phase extraction (SPE) packing material was then evaluated for extraction and clean-up of these bisphenols (BPs) from human urine samples. An accurate and sensitive analytical method based on the MIPMS-SPE coupled with HPLC-DAD has been successfully established for simultaneous determination of eight BPs from human urine samples with detection limits of 1.2–2.2 ng mL{sup −1}. The recoveries of BPs for urine samples at two spiking levels (100 and 500 ng mL{sup −1} for each BP) were in the range of 81.3–106.7% with RSD values below 8.3%.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1995

International Nuclear Information System (INIS)

1996-06-01

This report is the Atomic Energy Control Board (AECB) staff assessment of safety at the Pickering Nuclear Generating Station (PNGS-A and PNGS-B) for 1995. Our on-site Project Officers and Ottawa-based specialists monitored the stations throughout the year. In 1995, compliance with the Transportation Packaging of Radioactive Materials Regulations and the Cost Recovery Fees Regulations was satisfactory. The performance of the special safety systems was good. Releases of radioactive materials from the station were low and well below the legal limits for public safety. 10 tabs., 7 figs

Training and research on the nuclear reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

1998-01-01

The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

Modulation of Cyclodextrin Particle Amphiphilic Properties to Stabilize Pickering Emulsion.

Science.gov (United States)

Xi, Yongkang; Luo, Zhigang; Lu, Xuanxuan; Peng, Xichun

2018-01-10

Cyclodextrins have been proven to form complexes with linear oil molecules and stabilize emulsions. Amphiphilic properties of cyclodextrin particles were modulated through esterification reaction between β-cyclodextrin (β-CD) and octadecenyl succinic anhydride (ODSA) under alkaline conditions. ODS-β-CD particles with degree of substitution (DS) of 0.003, 0.011, and 0.019 were obtained. The introduced hydrophobic long chain that was linked within β-CD cavity led to the change of ODS-β-CD in terms of morphological structure, surface charge density, size, and contact angle, upon which the properties and stability of the emulsions stabilized by ODS-β-CD were highly dependent. The average diameter of ODS-β-CD particles ranged from 449 to 1484 nm. With the DS increased from 0.003 to 0.019, the contact angle and absolute zeta potential value of these ODS-β-CD particles improved from 25.7° to 47.3° and 48.1 to 62.8 mV, respectively. The cage structure of β-CD crystals was transformed to channel structure, then further to amorphous structure after introduction of the octadecenyl succinylation chain. ODS-β-CD particles exhibited higher emulsifying ability compared to β-CD. The resulting Pickering emulsions formed by ODS-β-CD particles were more stable during storage. This study investigates the ability of these ODS-β-CD particles to stabilize oil-in-water emulsions with respect to their amphiphilic character and structural properties.

Evaluation of severe accident risk in the Pickering a risk assessment

International Nuclear Information System (INIS)

Dinnie, K.S.; Raina, V.M.

1997-01-01

The nature of the design of commercial power plants is such that significant impacts on public health can only occur if a number of barriers fail. Rigorous design and licensing requirements ensure that the more likely accidents do not fail all these barriers and their contribution to risk is likely to be small. The task of estimating accident risk must, therefore, focus more towards those less likely but potentially more serious combinations of failures that are characterized by the following: a) a large release of fission products into the containment atmosphere, b) a breach in the containment envelope, and c) the existence of a driving force to expel the containment atmosphere to the outside environment. The likelihood of such conditions existing simultaneously during the course of an accident is expected to be small, such that experience and data regarding the behaviour of plant systems under such conditions is sparse or non-existent. The challenge of Probabilistic Safety Assessments (PSAs) is to examine the potential for severe accidents using approaches that are sufficiently detailed and realistic to provide valid information regarding plant risk and susceptibilities, while simple enough to keep the analysis manageable. This paper outlines the key features of the Pickering A Risk Assessment (PARA) (1) and the manner in which it addresses these issues, and provides some insights into the results and conclusions drawn from the study. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Otra Otra, F; Leira Rey, G

1971-07-01

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

IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

Modernization and Refurbishment of the RECH-1 Reactor

Energy Technology Data Exchange (ETDEWEB)

Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

2014-08-15

The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

Extensive utilisation of VR-1 reactor for nuclear education and training

International Nuclear Information System (INIS)

Rataj, J.

2010-01-01

The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

Use of the VR-1 ''Vrabec'' training reactor

International Nuclear Information System (INIS)

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

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

A study of wet deposition of atmospheric tritium releases at the Ontario Power Generation, Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Crooks, G.; DeWilde, J.; Yu, L.

2001-01-01

The Ontario Power Generation,Pickering Nuclear Generating Station (PNGS) has been investigating deposition of atmospheric releases of tritium on their site. This study has included numerical dispersion modelling studies conducted over the past three years, as well as an ongoing field monitoring study. The following paper will present results of the field monitoring study and make comparisons to the numerical modelling. The results of this study could be of potential use to nuclear stations in quantifying tritium deposition in near field regions where building wake effects dominate pollutant dispersion

Leakage rate from LOCA-aged inflatable airlock seals Pickering NGS 'B' personnel doors

International Nuclear Information System (INIS)

Fayle, G.W.; Cordingley, D.C.

1985-01-01

In order to demonstrate to the Atomic Energy Control Board that an air-lock inflatable seal will function after a LOCA exposure, an inflatable seal intended for personnel doors at the Pickering NGS 'B' was exposed to the thermal/moisture conditions of the LOCA requirement. While attending to determine the post-LOCA leakage rate it was found that additional leaks developed during each post-LOCA inflation/deflation cycle. The seal had been significantly and irreparably deteriorated by the LOCA exposure. The test has demonstrated that this type of LOCA exposed seal should not be expected to withstand either additional pressure above 207 kPa or additional inflation/deflation cycling. A higher inflation pressure and/or cycling will reduce the likelihood of a post-LOCA seal retaining an inflation pressure sufficient to prevent leakage across the seal

Evolution of CANDU vacuum building and pressure relief structures from Pickering NGS A to Darlington NGS A

International Nuclear Information System (INIS)

Beg, Z.M.; Ghosh, R.S.

1987-01-01

The vacuum building (VB) and pressure relief structures (PRS) are the unique features of multiple unit CANDU containments. In case of loss-of-coolant accident, the released radionuclides are drawn through the PRS into the subatmospheric VB, doused and contained without being released to the environment. This paper describes the differences in design, configuration and layout of the VB and PRS from Pickering NGS A to Darlington NGS A due to new developments in design concepts and to requirements which have proceeded from the experience gained in both the design and operation of the nuclear stations. (orig.)

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Operation and maintenance of 1MW PUSPATI TRIGA reactor

International Nuclear Information System (INIS)

Adnan Bokhari; Mohammad Suhaimi Kassim

2006-01-01

The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

International Nuclear Information System (INIS)

Frescura, G.M.; Smith, A.J.; Lau, J.H.

1991-01-01

Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

Perspectives on reactor safety. Revision 1

International Nuclear Information System (INIS)

Haskin, F.E.; Hodge, S.A.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

Perspectives on reactor safety. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Production of starch nanoparticles by dissolution and non-solvent precipitation for use in food-grade Pickering emulsions.

Science.gov (United States)

Saari, Hisfazilah; Fuentes, Catalina; Sjöö, Malin; Rayner, Marilyn; Wahlgren, Marie

2017-02-10

The aim of this study was to investigate non-solvent precipitation of starch to produce nanoparticles that could be used in Pickering emulsions. The material used was waxy maize, modified with octenyl succinic anhydride. Different methods of non-solvent precipitation were investigated, and a method based on direct mixing of an 8% starch solution and ethanol (ratio 1:1) was found to produce the smallest particles. The particle size was measured using AFM and AF4, and was found to be in the range 100-200nm. However, both larger particles and aggregates of nanoparticles were observed. The emulsion produced using the precipitated starch particles had a droplet size that between 0.5 and 45μm, compared to emulsions produced from waxy maize granules, in which had a size of 10-100μm. The drop in size contributed to increased stability against creaming. The amount of starch used for emulsion stabilization could also be substantially reduced. Copyright © 2016 The Author(s). Published by Elsevier Ltd.. All rights reserved.

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Advances in Reactor Physics, Mathematics and Computation. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

International Nuclear Information System (INIS)

Perry, E.F.

1995-01-01

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

Reactor utilization, Part 1

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1981-01-01

The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

International Nuclear Information System (INIS)

Salahuddin, A.; Israr, M.; Hussain, M.

2002-01-01

Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

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

1985-01-01

According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

Energy Technology Data Exchange (ETDEWEB)

Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

2013-07-01

Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Interpretation of 131I hippuran renocystogram using vascular invasion segment systemic flow, and DYNA CAMERA II Picker

International Nuclear Information System (INIS)

Morcellet, J.L.; Baret, A.

A quantitative approximation of flows of fluids from each kidney (renal clearances urinary flows), of the hippuran mean stay time into each kidney was proposed. These times are decomposed into cortical transit mean time and into pyelocavities mean stay time. The use of a dual isotope scintillation Dyna Camera II Picker changes the collecting of the data and permits the simultaneous measurement of cardiac output which is required for their treatment. This treatment is carried out by the mean of a videotape recorder which authorizes delayed time work and by the mean of a hundred channels computer, which displays numerical data and their integration [fr

New human machine interface for VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

2002-01-01

The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

International Nuclear Information System (INIS)

1988-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

Department of Reactor Technology annual progress report 1 January -31 December 1977

International Nuclear Information System (INIS)

1978-04-01

The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

Preparation of cellulose nanocrystals from asparagus (Asparagus officinalis L.) and their applications to palm oil/water Pickering emulsion.

Science.gov (United States)

Wang, Wenhang; Du, Guanhua; Li, Cong; Zhang, Hongjie; Long, Yunduo; Ni, Yonghao

2016-10-20

Nano cellulosic materials as promising emulsion stabilizers have attracted great interest in food industry. In this paper, five different sized cellulose nanocrystals (CNC) samples were prepared from stem of Asparagus officinalis L. using the same sulfuric acid hydrolysis conditions but different times (1.5, 2, 2.5, 3.0, and 3.5h). The sizes of these CNC ranged from 178.2 to 261.8nm, with their crystallinity of 72.4-77.2%. The CNC aqueous dispersions showed a typical shear thinning behavior. In a palm oil/water (30/70, v/v) model solution, stable Pickering emulsions were formed with the addition of CNC, and their sizes are in the range of 1-10μm based on the optical and confocal laser scanning microscopy (CLSM) observation. The CNC sample prepared at 3h hydrolysis time, showed a relative efficient emulsion capacity for palm oil droplets, among these CNCs. Other parameters including the CNC, salt, and casein concentrations on the emulsion stability were studied. Copyright © 2016 Elsevier Ltd. All rights reserved.

Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

International Nuclear Information System (INIS)

1989-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

International Nuclear Information System (INIS)

Ratanatongchai, Wichian

2009-07-01

Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

New measuring and protection system at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Jurickova, M.

2006-01-01

The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

International Nuclear Information System (INIS)

1991-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

International Nuclear Information System (INIS)

1992-08-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Recovery of tritium from CANDU reactors, its storage and monitoring of its migration in the environment

International Nuclear Information System (INIS)

Holtslander, W.J.; Osborne, R.V.

1979-07-01

Tritium is produced in CANDU heavy water reactors mainly by neutron activation of deuterium. The typical production rate is 2.4 kCi per megawatt-year (89 TBq. per megawatt-year. In Pickering Generating Station the average concentration of tritium in the moderators has reached 16 Ci.kg -1 (0.6 TBq.kg -1 ) and in coolants, 0.5 Ci.kg -1 (0.02 TBq.kg -1 ). Concentrations will continue to increase towards an equilibrium determined by the production rate, the tritium decay rate and heavy water replacement. Tritium removal methods that are being considered for a pilot plant design are catalytic exchange of DTO with D 2 and electrolysis of D 2 O/DTO to provide feed for cryogenic distillation of D 2 /DT/T 2 . Storage methods for the removed tritium - as elemental gas, as metal hydrides and in cements - are also being investigated. Transport of tritiated wastes should not be a particularly difficult problem in light of extensive experience in transporting tritiated heavy water. Methods for determining the presence of tritium in the environment of any tritium handling facility are well established and have the capability of measuring concentrations of tritium down to current ambient values. (author)

Quinoa starch granules as stabilizing particles for production of Pickering emulsions.

Science.gov (United States)

Rayner, Marilyn; Sjöö, Malin; Timgren, Anna; Dejmek, Petr

2012-01-01

Intact starch granules isolated from quinoa (Chenopodium quinoa Willd.) were used to stabilize emulsion drops in so-called Pickering emulsions. Miglyol 812 was used as dispersed phase and a phosphate buffer (pH7) with different salt (NaCl) concentrations was used as the continuous phase. The starch granules were hydrophobically modified to different degrees by octenyl succinic anhydride (OSA) or by dry heat treatment at 120 degrees C in order to study the effect on the resulting emulsion drop size. The degree of OSA-modification had a low to moderate impact on drop size. The highest level of modification (4.66%) showed the largest mean drop size, and lowest amount of free starch, which could be an effect of a higher degree of aggregation of the starch granules and, thereby, also the emulsion drops stabilized by them. The heat treated starch granules had a poor stabilizing ability and only the starch heated for the longest time (150 min at 120 degrees C) had a better emulsifying capacity than the un-modified native starch granules. The effect of salt concentration was rather limited. However, an increased concentration of salt slightly increased the mean drop size and the elastic modulus.

Facile Route to Transparent, Strong, and Thermally Stable Nanocellulose/Polymer Nanocomposites from an Aqueous Pickering Emulsion.

Science.gov (United States)

Fujisawa, Shuji; Togawa, Eiji; Kuroda, Katsushi

2017-01-09

Cellulose nanofibril (CNF) is a promising nanofiller for polymer nanocomposite materials, and a critical challenge in designing these materials is organization of the nanostructure using a facile process. Here, we report a facile aqueous preparation process for nanostructured polystyrene (PS)/CNF composites via the formation of a CNF-stabilized Pickering emulsion. PS nanoparticles, with a narrow size distribution, were synthesized by free radical polymerization in water using CNF as a stabilizer. The nanoparticles were easily collected by filtration, and the resulting material had a composite structure of PS nanoparticles embedded in a CNF framework. The PS/CNF nanocomposite showed high optical transparency, strength, and thermal dimensional stability. Thus, this technique provides a simple and environmentally friendly method for the preparation of novel CNF/polymer nanocomposite materials.

Thermohydraulic analysis for power increase of IEAR-1 reactor

International Nuclear Information System (INIS)

Umbehaun, Pedro E.; Bastos, Jose L.F.

1996-01-01

In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

1981-12-01

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

Anemia em catadores de material reciclável que utilizam carrinho de propulsão humana no município de Santos Anemia in recyclable waste pickers using human driven pushcarts in the city of Santos, southeastern Brazil

Directory of Open Access Journals (Sweden)

Mauro Abrahão Rozman

2010-06-01

Full Text Available OBJETIVO: Estimar a prevalência de anemia e analisar os fatores de risco a ela associados nos catadores de material reciclável que utilizam carrinho de propulsão humana do município de Santos - São Paulo. MÉTODO: Estudo transversal com 253 catadores foi realizado em julho de 2005. A coleta de informações foi feita por meio de questionário com informações sobre características individuais, ocupacionais e dietéticas. Foi realizada avaliação antropométrica e coletado sangue venoso para hemograma completo e sorologias de HIV, HCV, HBV e sífilis. A análise estatística foi feita por análise uni e multivariada (regressão logística, relacionando a anemia aos fatores de risco. RESULTADOS: A prevalência de anemia foi de 38,3%. As variáveis que mostraram associação independente com anemia no modelo multivariado foram: sexo (OR 4,11; IC95%: 1,56-10,87, infecção pelo HIV (OR 9,23; IC95%: 2,93-29,1, IMC (OR 0,21; IC95%: 0,07-0,64, anos de trabalho como catador (OR 4,54; IC95%: 1,29-16,0, consumo de leite (OR 0,36; IC95%: 0,16-0,81 e de proteína animal (OR 0,39; IC95%: 0,15-0,97. CONCLUSÃO: A prevalência de anemia entre catadores de material reciclável é elevada mesmo após a obrigatoriedade de adição de ferro nas farinhas de trigo e milho. Os catadores são excluídos das ações de proteção à saúde do trabalhador, previstas na legislação. Ações de saúde dirigidas a essa categoria profissional devem ser implementadas, garantindo a acessibilidade aos serviços de saúde.OBJECTIVE: To assess the prevalence of anemia and describe associated risk factors in recyclable waste pickers using human-driven pushcarts in the city of Santos. METHODS: A cross-sectional study including 253 recyclable waste pickers was conducted in the city of Santos, southeastern Brazil, in July 2005. A questionnaire was used to collect information about individual, occupational, and dietary factors. All subjects underwent an anthropometric

Department of Reactor Technology: annual progress report 1 January - 31 December 1976

International Nuclear Information System (INIS)

1977-06-01

The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Neutron density optimal control of A-1 reactor analoque model

International Nuclear Information System (INIS)

Grof, V.

1975-01-01

Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

Department of Reactor Technology annual progress report 1 January - 31 December 1978

International Nuclear Information System (INIS)

1979-04-01

The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

International Nuclear Information System (INIS)

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1993-March 31, 1994

International Nuclear Information System (INIS)

1994-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

International Nuclear Information System (INIS)

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Modification of the IAN-R1 reactor

International Nuclear Information System (INIS)

Jaime, J.; Ahumada, S.; Spin, R.A.

1990-01-01

The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

THE INFLUENCE OF TRUST IN THE CONSTITUTION OF A BRAZILIAN COOPERATIVE OF SELECTIVE WASTE PICKERS

Directory of Open Access Journals (Sweden)

Dayanne Marciane Gonçalves

2016-03-01

Full Text Available Since the Brazilian public policy started to encourage solidarity economy in 2003, the number of projects and enterprises in this sector has steadily increased. Embeddedness has contributed to the understanding of organizational phenomena of solidarity economy and cooperatives. The aim of this study was to understand the influence of trust, from the perspective of Mark Granovetter’s social networks, on the constitution of a cooperative of urban recyclable waste pickers in southern Brazil between 1996 until early 2012, considered the foundation period. We used the qualitative method with a historical approach to social relationships and content analysis. Possible influences of trust were analysed based on the economic, social and political history of the cooperative. Among the main results, we highlight the existence of social relations before the constitution, defined by trust due to family identity and reputation built over time.

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

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

International Nuclear Information System (INIS)

Jumel, Stephanie; Van-Duysen, Jean Claude

2005-01-01

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

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

International Nuclear Information System (INIS)

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

Energy Technology Data Exchange (ETDEWEB)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

International Nuclear Information System (INIS)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Present and future activities of TRIGA RC-1 Reactor

International Nuclear Information System (INIS)

Festinesi, A.

1986-01-01

A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

Neutronic studies in the enrichment reduction of research reactor IEAR-1

International Nuclear Information System (INIS)

Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

1987-01-01

In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

Fragger: a protein fragment picker for structural queries [version 2; referees: 2 approved

Directory of Open Access Journals (Sweden)

Francois Berenger

2018-04-01

Full Text Available Protein modeling and design activities often require querying the Protein Data Bank (PDB with a structural fragment, possibly containing gaps. For some applications, it is preferable to work on a specific subset of the PDB or with unpublished structures. These requirements, along with specific user needs, motivated the creation of a new software to manage and query 3D protein fragments. Fragger is a protein fragment picker that allows protein fragment databases to be created and queried. All fragment lengths are supported and any set of PDB files can be used to create a database. Fragger can efficiently search a fragment database with a query fragment and a distance threshold. Matching fragments are ranked by distance to the query. The query fragment can have structural gaps and the allowed amino acid sequences matching a query can be constrained via a regular expression of one-letter amino acid codes. Fragger also incorporates a tool to compute the backbone RMSD of one versus many fragments in high throughput. Fragger should be useful for protein design, loop grafting and related structural bioinformatics tasks.

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

Steps to Advanced CANDU 600

International Nuclear Information System (INIS)

Oh, Yongshick; Brooks, G. L.

1988-01-01

The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

Reliabitity study of the accumulator system for Angra-1 reactor

International Nuclear Information System (INIS)

Santos Maciel, C.C.R.

1980-01-01

The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

Design of a new research reactor : 1st year conceptual design

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T.

2004-01-01

A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

Education and research at the VR-1 Vrabec training reactor facility

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

Graphite stack corrosion of BUGEY-1 reactor (synthesis)

International Nuclear Information System (INIS)

Petit, A.; Brie, M.

1996-01-01

The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

Sandia reactor kinetics codes: SAK and PK1D

International Nuclear Information System (INIS)

Pickard, P.S.; Odom, J.P.

1978-01-01

The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

Design and installation of a strategically placed algae mesh barrier at OPG Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Marttila, D.; Patrick, P.; Gregoris, C.

2009-01-01

Ontario Power Generation's Pickering Nuclear has experienced a number of events in which attached algae have become entrained in the water intake costing approximately $30M over the 1995-2005 period as a result of deratings, Unit shutdowns and other operational issues. In 2005-2006 OPG and Kinectrics worked collaboratively on evaluating different potential solutions to reduce the impact of algae on the station. One of the solutions developed by Kinectrics included a strategically placed barrier net designed to regulate algae flow into the station intake. In 2006, Kinectrics designed and installed the system, the first of its kind at a Nuclear Power Plant in Canada. The system was operational by May 2007. OPG completed an effectiveness study in 2007 and concluded the barrier system had a beneficial effect on reducing algae impact on the station. (author)

Ageing problems and renovation programme of ET-RR-1 reactor

International Nuclear Information System (INIS)

Khattab, M.S.; Sultan, M.A.

1995-01-01

Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

Generating the flux map of Nigeria Research Reactor-1 for efficient ...

African Journals Online (AJOL)

One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

Reactor protection system. Revision 1

International Nuclear Information System (INIS)

Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Digital computer control on Canadian nuclear power plants -experience to date and the future outlook

International Nuclear Information System (INIS)

Pearson, A.

1977-10-01

This paper discusses the performance of the digital computer control system at Pickering through the years 1973 to 1976. This evaluation is based on a study of the Pickering Generating Station operating records. The paper goes on to explore future computer architectures and the advantages that could accrue from a distributed system approach. Also outlined are the steps being taken to develop these ideas further in the context of two Chalk River projects - REDNET, an advanced data acquisition system being installed to process information from engineering experiments in NRX and NRU reactors, and CRIP, a prototype communications network using cable television technology. (author)

Radiological pathways analysis for spent solvents from the boiler chemical cleaning at the Pickering Nuclear Site

International Nuclear Information System (INIS)

Garisto, N.C.; Eslami, Z.; Hodgins, S.; Beaman, T.; Von Svoboda, S.; Marczak, J.

2006-01-01

Spent solvents are generated as a result of Boiler Chemical Cleanings (BCC) at CANDU reactor sites. These solutions contain small amount of radioactivity from a number of different sources including: Cut tubes - short sections of boiler tubes are infrequently removed from the boilers for a detailed characterization. These tubes are typically only plugged at the tubesheet allowing the primary side deposits to be exposed to BCC solvents. Tube leaks - primary to secondary side leaks also occur infrequently as a result of tube degradation. Radioactivity from the leaking fluid can consequently be deposited in the sludge on the secondary side of the tubes. Diffusion of tritium - during normal operation of the reactor units, tritium slowly diffuses from the heavy water in the primary heat-transfer system to the light-water coolant on the secondary side. Some of this tritium is retained in the secondary side deposits. The Pickering Nuclear Generating Station (PNGS) would like the flexibility to have several options for handling the spent solvent waste and associated rinse water from BCC. To this end, a radiological pathways analysis was undertaken to determine dose consequences associated with each option. Sample results from this study are included in this paper. The pathways analysis is used in this study to calculate dose to hypothetical receptors including individuals such as truck drivers, incinerator workers, residue (ash) handlers, residents who live near the landfill, inadvertent intruders into the landfill after closure and residents who live near the outfall. This dose is compared to a de minimis dose. A de minimis dose or dose rate represents a level of risk, which is generally accepted as being of no significance. Shipments of spent solvents and rinse water with corresponding doses below de minimis can be sent to conventional (i.e., non-radioactive) landfills for incineration and disposal as the radioactive dose associated with them is much less than natural

Safety in nuclear power systems

International Nuclear Information System (INIS)

Myers, L.C.

1987-05-01

This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents that have occurred to date. Details are also provided of Ontario Hydro's problems with Unit 2 at Pickering

Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

International Nuclear Information System (INIS)

Kuruc, J.; Matel, L.

2007-01-01

The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

Demolition of the FRJ-1 research reactor (MERLIN)

International Nuclear Information System (INIS)

Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2003-01-01

FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

Evaporation rate measurement in the pool of IEAR-1 reactor

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

2000-01-01

The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

Energy Technology Data Exchange (ETDEWEB)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

RA reactor operation and maintenance in 1992, Part 1

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

1992-01-01

During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

Spent fuel management - two alternatives at the FiR 1 reactor

International Nuclear Information System (INIS)

Salmenhaara, S.E.J.

2001-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

Computer-aided testing and operational aids for PARR-1 nuclear reactor

International Nuclear Information System (INIS)

Ansari, S.A.

1990-01-01

The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

Properties of autoregressive model in reactor noise analysis, 1

International Nuclear Information System (INIS)

Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

1987-01-01

Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

Spent fuel management plans for the FiR 1 Reactor

International Nuclear Information System (INIS)

Salmenhaara, S. E. J.

2002-01-01

The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

Experimental study of the IPR-R1 TRIGA reactor power channels responses

International Nuclear Information System (INIS)

Mesquita, Henrique F.A.; Ferreira, Andrea V.

2015-01-01

The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

Irradiation experience of IPEN fuel at IEA-R1 research reactor

International Nuclear Information System (INIS)

Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

RA reactor operation and maintenance in 1996, Part 1

International Nuclear Information System (INIS)

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

1996-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

Possible future roles for fast breeder reactors Part 1 and 2

International Nuclear Information System (INIS)

1978-06-01

Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

The AMPS 1.5 MW low-pressure compact reactor

International Nuclear Information System (INIS)

Hewitt, J.S.

1987-01-01

The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

Flux distribution measurements in the Bruce A unit 1 reactor

International Nuclear Information System (INIS)

Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

1977-07-01

Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2002-07-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

Science.gov (United States)

El-Genk, Mohamed S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

International Nuclear Information System (INIS)

El-genk, M.S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

Suitability of CR-39 dosimeters for personal dosimetry around CANDU reactors

International Nuclear Information System (INIS)

Cross, W.G.

1992-08-01

The capabilities and limitations of CR-39 damage track detectors have been evaluated for their use as personal neutron dosimeters around CANDU reactors. Since the energy response is a critical characteristic, the neutron energy spectra expected within CANDU containments were studied. In the boiler rooms, around the moderator cooling systems, and in most of the fueling machine vaults, the spectra vary considerably, but the majority of the dose is expected to be delivered by neutrons above 80 keV, the approximate threshold for electrochemically-etched CR-39 detectors. In the Pickering A fueling machine vault, and in areas in other stations to which neutrons from reactors have been multiply scattered, lower energy neutrons may be important. In nearly all areas where people work, it appears that working times will be limited by gamma rays rather than by neutrons. The characteristics of other neutron dosimeters - bubble and superheated drop detectors, albedo detectors, and Si real-time detectors - were also reviewed. For workers who typically receive neutron doses that are small compared with regulatory limits, CR-39 is the most suitable available dosimeter for demonstrating compliance. All single dosimeters have poor angular response over the range 0 to 180 degrees because of the shielding of the body. Albedo and Si detectors have particularly poor energy responses over the energy range of importance. Bubble and superheated drop detectors have the advantages of immediate readout and high sensitivity, but the disadvantages of inability to integrate doses over a long period, temperature dependence, very limited range and higher cost. (Author) (110 refs., 45 figs.)

CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1); CAC-RA-1 1958-1998. Los primeros anios del CAC. Historia del primer reactor nuclear argentino (RA-1)

Energy Technology Data Exchange (ETDEWEB)

Forlerer, Elena; Palacios, Tulio A [comps.

1998-07-01

After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation.

Benchmark testing of Canadol-2.1 for heavy water reactor

International Nuclear Information System (INIS)

Liu Ping

1999-01-01

The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

Energy Technology Data Exchange (ETDEWEB)

Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

2017-05-15

Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

Energy Technology Data Exchange (ETDEWEB)

Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

1999-08-01

The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

International Nuclear Information System (INIS)

Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

1999-01-01

The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

Reactor inventory monitoring system for Angra-1 reactor

International Nuclear Information System (INIS)

S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

1996-01-01

This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

Modifications done in the IPR-R1 reactor and their auxiliary systems

International Nuclear Information System (INIS)

Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

1986-01-01

The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

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

Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

International Nuclear Information System (INIS)

Blom, K.H.; Falck, K.; Krull, W.

1990-01-01

The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

Instrumentation renewal at the FIR 1 research reactor in Finland

International Nuclear Information System (INIS)

Bars, Bruno; Kall, Leif

1982-01-01

The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

Dose measurements in controlled area of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Alvarenga, F.L.; Junior, F.M.

2005-01-01

The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

Energy Technology Data Exchange (ETDEWEB)

Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

2003-06-01

FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

Equipment for neutron measurements at VR-1 Sparrow training reactor

International Nuclear Information System (INIS)

Kolros, Antonin; Huml, Ondrej; Kos, Josef

2008-01-01

Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

International Nuclear Information System (INIS)

Olsson, S.; Andermo, L.

1980-01-01

This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

International Nuclear Information System (INIS)

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

1986-01-01

In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

Production of Concentrated Pickering Emulsions with Narrow Size Distributions Using Stirred Cell Membrane Emulsification.

Science.gov (United States)

Manga, Mohamed S; York, David W

2017-09-12

Stirred cell membrane emulsification (SCME) has been employed to prepare concentrated Pickering oil in water emulsions solely stabilized by fumed silica nanoparticles. The optimal conditions under which highly stable and low-polydispersity concentrated emulsions using the SCME approach are highlighted. Optimization of the oil flux rates and the paddle stirrer speeds are critical to achieving control over the droplet size and size distribution. Investigating the influence of oil volume fraction highlights the criticality of the initial particle loading in the continuous phase on the final droplet size and polydispersity. At a particle loading of 4 wt %, both the droplet size and polydispersity increase with increasing of the oil volume fraction above 50%. As more interfacial area is produced, the number of particles available in the continuous phase diminishes, and coincidently a reduction in the kinetics of particle adsorption to the interface resulting in larger polydisperse droplets occurs. Increasing the particle loading to 10 wt % leads to significant improvements in both size and polydispersity with oil volume fractions as high as 70% produced with coefficient of variation values as low as ∼30% compared to ∼75% using conventional homogenization techniques.

Electrospun composite matrices of poly(ε-caprolactone)-montmorillonite made using tenside free Pickering emulsions.

Science.gov (United States)

Samanta, Archana; Takkar, Sonam; Kulshreshtha, Ritu; Nandan, Bhanu; Srivastava, Rajiv K

2016-12-01

The production of composite electrospun matrices of poly(ε-caprolactone) (PCL) using an emulsifier-free emulsion, made with minimal organic solvent, as precursor is reported. Pickering emulsions of PCL were prepared using modified montmorillonite (MMT) clay as the stabilizer. Hydrophobic tallow group of the modified MMT clay resulted in analogous interaction of clay with oil and aqueous phase and its adsorption at the interface to provide stability to the resultant emulsion. Composite fibrous matrices of PCL and MMT were produced using electrospinning under controlled conditions. The fiber fineness was found to alter with PCL concentration and volume fraction of the aqueous and oil phases. A higher tensile strength and modulus was obtained with inclusion of MMT in PCL electrospun matrix in comparison to a matrix made using neat PCL. The presence of clay in the fibrous matrix did not change the cell proliferation efficiency in comparison to neat PCL matrix. Composite fibrous matrices of PCL/MMT bearing enhanced tensile properties may find applications in areas other than tissue engineering for example food packaging and filtration. Copyright © 2016 Elsevier B.V. All rights reserved.

Investigations of titamium and zirconium hydrides to determine suitability of recoverable tritium immobilization for the Pickering tritium removal system

International Nuclear Information System (INIS)

Noga, J.O.

1981-11-01

A tritium removal system will be constructed at Pickering Nuclear Generating station to reduce the adverse effects of this radioactive hydrogen isotope. This report summarizes various properties of titanium and zirconium sponge hydrides which have been selected as suitable candidates for tritium product immobilization. Equilibrium pressure-composition-temperature data indicates that both materials behave suitably to provide a safe, solid form of tritium storage. Titanium tritide is recommended as the best choice due to higher dissociation pressures which can be achieved at equivalent temperatures when compared to zirconium tritide. Higher dissociation pressures would result in faster and more efficient recovery of tritium gas from the immobilized state. It is evident from the stability of these compounds that their utilization as tritides will greatly enhance the integrity of tritium storage

Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

2006-01-01

The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

Recent fuel handling experience in Canada

International Nuclear Information System (INIS)

Welch, A.C.

1991-01-01

For many years, good operation of the fuel handling system at Ontario Hydro's nuclear stations has been taken for granted with the unavailability of the station arising from fuel handling system-related problems usually contributing less than one percent of the total unavailability of the stations. While the situation at the newer Hydro stations continues generally to be good (with the specific exception of some units at Pickering B) some specific and some general problems have caused significant loss of availability at the older plants (Pickering A and Bruce A). Generally the experience at the 600 MWe units in Canada has also continued to be good with Point Lepreau leading the world in availability. As a result of working to correct identified deficiencies, there were some changes for the better as some items of equipment that were a chronic source of trouble were replaced with improved components. In addition, the fuel handling system has been used three times as a delivery system for large-scale non destructive examination of the pressure tubes, twice at Bruce and once at Pickering and performing these inspections this way has saved many days of reactor downtime. Under COG there are several programs to develop improved versions of some of the main assemblies of the fuelling machine head. This paper will generally cover the events relating to Pickering in more detail but will describe the problems with the Bruce Fuelling Machine Bridges since the 600 MW 1P stations have a bridge drive arrangement that is somewhat similar to Bruce

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

International Nuclear Information System (INIS)

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

1988-01-01

According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

IPR-R1 TRIGA research reactor decommissioning plan

International Nuclear Information System (INIS)

Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

2008-01-01

The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

The Chernobyl reactor accident. Pt. 1 and 2

International Nuclear Information System (INIS)

1986-06-01

The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

Fabrication and evaluation of chitosan/NaYF4:Yb3+/Tm3+ upconversion nanoparticles composite beads based on the gelling of Pickering emulsion droplets

International Nuclear Information System (INIS)

Yan, Huiqiong; Chen, Xiuqiong; Shi, Jia; Shi, Zaifeng; Sun, Wei; Lin, Qiang; Wang, Xianghui; Dai, Zihao

2017-01-01

The rare earth ion doped upconversion nanoparticles (UCNPs) synthesized by hydrophobic organic ligands possess poor solubility and low fluorescence quantum yield in aqueous media. To conquer this issue, NaYF 4 :Yb 3+ /Tm 3+ UCNPs, synthesized by a hydrothermal method, were coated with F127 and then assembled with chitosan to fabricate the chitosan/NaYF 4 :Yb 3+ /Tm 3+ composite beads (CS/NaYF 4 :Yb 3+ /Tm 3+ CBs) by Pickering emulsion system. The characterization results revealed that the as-synthesized NaYF 4 :Yb 3+ /Tm 3+ UCNPs with an average size of 20 nm exhibited spherical morphology, high crystallinity and characteristic emission upconversion fluorescence with an overall blue color output. The NaYF 4 :Yb 3+ /Tm 3+ UCNPs were successfully conjugated on the surface of chitosan beads by the gelling of emulsion droplets. The resultant CS/NaYF 4 :Yb 3+ /Tm 3+ CBs showed good upconversion luminescent property, drug-loading capacity, release performance and excellent biocompatibility, exhibiting great potentials in targeted drug delivery and tissue engineering with potential tracking capability and lasting release performance. - Highlights: • NaYF 4 :Yb 3+ /Tm 3+ UCNPs were coated by F127 to improve aqueous dispersibility. • NaYF 4 :Yb 3+ /Tm 3+ UCNPs were assembled with chitosan to fabricate the composite beads (CMs). • Pickering emulsions stabilized by UCNPs exhibited uniform and satisfactory emulsion droplets. • The CMs prepared by the gelling of emulsion droplet preserved upconversion luminescent property. • The resultant CMs showed good drug-loading capacity, release performance and biocompatibility.

Current activities at the FiR 1 TRIGA reactor

International Nuclear Information System (INIS)

Salmenhaara, Seppo

2002-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Expanding the storage capability at ET-RR-1 research reactor at Inshass

International Nuclear Information System (INIS)

Sultan, Mariy M.; Khattab, M.

1999-01-01

Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

Licensing of the first reload of Angra-1 reactor

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1985-01-01

The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

Evaluation of the trial design studies for an advanced marine reactor, (1)

International Nuclear Information System (INIS)

1988-03-01

The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

Transformation of 1,1,1-trichloroethane in an anaerobic packed-bed reactor at various concentrations of 1,1,1-trichloroethane, acetate and sulfate

NARCIS (Netherlands)

deBest, JH; Jongema, H; Weijling, A; Doddema, HJ; Janssen, DB; Harder, W

Biotransformation of 1,1,1-trichloroethane (CH3CCl3) was observed in an anaerobic packed-bed reactor under conditions of both sulfate reduction and methanogenesis. Acetate (1 mM) served as an electron donor. CH3CCl3 was completely converted up to the highest investigated concentration of 10 mu M.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

2014-08-15

The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1985-01-01

This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

International Nuclear Information System (INIS)

Andermo, L.; Sundman, B.

1974-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

New digital control system for the operation of the Colombian research reactor IAN-R1

International Nuclear Information System (INIS)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

2015-09-01

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias

2005-01-01

The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

Validation of moderator-level reactivity coefficient using station data

Energy Technology Data Exchange (ETDEWEB)

Younis, M.; Martchouk, I., E-mail: mohamed.younis@amecfw.com, E-mail: iouri.martchouk@amecfw.com [Amec Foster Wheeler, Toronto, ON (Canada); Buchan, P.D., E-mail: david.buchan@opg.com [Ontario Power Generation, Pickering, ON (Canada)

2015-07-01

The reactivity effect due to variations in the moderator level has been recognized as a reactor physics phenomenon of importance during normal operation and accident analysis. The moderator-level reactivity coefficient is an important parameter in safety analysis of CANDU reactors, e.g., during Loss of Moderator Heat Sink as well as in the simulation of Reactor Regulating System action in CANDU reactors that use moderator level for reactivity control. This paper presents the results of the validation exercise of the reactor-physics toolset using the measurements performed in Pickering Unit 4 in 2003. The capability of the code suite of predicting moderator-level reactivity effect was tested by comparing measured and predicted reactor-physics parameters. (author)

Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

International Nuclear Information System (INIS)

Andermo, L.

1977-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

Energy Technology Data Exchange (ETDEWEB)

El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

International Nuclear Information System (INIS)

El-Messeiry, A.M.

1996-01-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

International Nuclear Information System (INIS)

Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

1964-01-01

1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

Energy Technology Data Exchange (ETDEWEB)

Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2007-12-15

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

Preparation of Lignin/Sodium Dodecyl Sulfate Composite Nanoparticles and Their Application in Pickering Emulsion Template-Based Microencapsulation.

Science.gov (United States)

Pang, Yuxia; Wang, Shengwen; Qiu, Xueqing; Luo, Yanling; Lou, Hongming; Huang, Jinhao

2017-12-20

Lignin is a vastly underutilized biomass resource. The preparation of water-dispersed lignin nanoparticles is an effective way to realize the high-value utilization of lignin. However, the currently reported preparation methods of lignin nanoparticles still have some drawbacks, such as the requirement for toxic organic solvent or chemical modification, complicated operation process, and poor dispersibility. Here, lignin/sodium dodecyl sulfate (SDS) composite nanoparticles (LSNPs) with outstanding water dispersibility and a size range of 70-200 nm were facilely prepared via acidifying the mixed basic solution of alkaline lignin and SDS. No harsh chemical was needed. The formation mechanism was systematically studied. Results indicated that the LSNPs were obtained by acid precipitation of the mixed micelles formed by the self-assembly of lignin and SDS. In addition, on the basis of the LSNP-stabilized Pickering emulsions, lignin/polyurea composite microcapsules combining the excellent chemical stability of a synthetic polyurea shell with the fantastic antiphotolysis and antioxidant properties of lignin were successfully prepared.

Automated phase picker and source location algorithm for local distances using a single three component seismic station

International Nuclear Information System (INIS)

Saari, J.

1989-12-01

The paper describes procedures for automatic location of local events by using single-site, three-component (3c) seismogram records. Epicentral distance is determined from the time difference between P- and S-onsets. For onset time estimates a special phase picker algorithm is introduced. Onset detection is accomplished by comparing short-term average with long-term average after multiplication of north, east and vertical components of recording. For epicentral distances up to 100 km, errors seldom exceed 5 km. The slowness vector, essentially the azimuth, is estimated independently by using the Christoffersson et al. (1988) 'polarization' technique, although a priori knowledge of the P-onset time gives the best results. Differences between 'true' and observed azimuths are generally less than 12 deg C. Practical examples are given by demonstrating the viability of the procedures for automated 3c seismogram analysis. The results obtained compare favourably with those achieved by a miniarray of three stations. (orig.)

Welding of the A1 reactor pressure vessel

International Nuclear Information System (INIS)

Becka, J.

1975-01-01

As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

Determination flux in the Reactor JEN-1

International Nuclear Information System (INIS)

Manas Diaz, L.; Montes Ponce de leon, J.

1960-01-01

This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Coles, G.A.; McKay, S.L.

1995-06-01

A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

Science.gov (United States)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

West, C D [comp.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

West, C.D.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

International Nuclear Information System (INIS)

1987-01-01

Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

International Nuclear Information System (INIS)

Tikhomirov, L.N.

1995-01-01

The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

RA reactor operation and maintenance in 1994, Part 1

International Nuclear Information System (INIS)

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

1994-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

Static yield stress of a magnetorheological fluid containing Pickering emulsion polymerized Fe2O3/polystyrene composite particles.

Science.gov (United States)

Seo, Youngwook P; Kwak, Soonjong; Choi, Hyoung Jin; Seo, Yongsok

2016-02-01

The flow behaviors of magnetorheological (MR) suspensions containing Pickering emulsion polymerized Fe2O3/polystyrene (PS) composite particles were reanalyzed using the Seo-Seo model. The experimental shear stress data obtained experimentally from the magnetorheological fluid fit well to the Seo-Seo model, indicating that this model can describe the structural reformation process of the aligned fibers at various shear rates. Unlike the dynamic yield stress obtained from the Cho-Choi-Jhon (CCJ) model, the static yield stresses obtained from the Seo-Seo model exhibit the same quadratic dependence on the magnetic field strength for both pure Fe2O3 particle suspension and Fe2O3/PS particle suspensions, which is in agreement with the predictions of the polarization model. The static yield stress plausibly explains the difference in underlying mechanism of MR fluids. Copyright © 2015 Elsevier Inc. All rights reserved.

Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

Manas Diaz, L; Montes Ponce de leon, J.

1960-07-01

This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

International Nuclear Information System (INIS)

Odoi, H.C.

2014-06-01

The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

New burnup calculation of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

2015-01-01

The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

International Nuclear Information System (INIS)

1986-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

International Nuclear Information System (INIS)

1978-05-01

An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

In situ sampling for pressure tube deuterium concentration

International Nuclear Information System (INIS)

Harrington, A.J.; Kittmer, C.A.

1988-01-01

The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

Problems of nuclear reactor safety. Vol. 1

International Nuclear Information System (INIS)

Shal'nov, A.V.

1995-01-01

Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

Auxiliary control system of the safety parameters for IPR-R1 reactor

International Nuclear Information System (INIS)

Coura, J.G.

1986-01-01

This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

OSCAR-4 Code System Application to the SAFARI-1 Reactor

International Nuclear Information System (INIS)

Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

2008-01-01

The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

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

International Nuclear Information System (INIS)

Novelli, A.

1982-01-01

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

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1973-07-01

By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

International Nuclear Information System (INIS)

1987-08-01

Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

International Nuclear Information System (INIS)

1982-01-01

A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

Modelling of the RA-1 reactor using a Monte Carlo code; Modelado del reactor RA-1 utilizando un codigo Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares

2000-07-01

It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

International Nuclear Information System (INIS)

Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

1976-01-01

The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

Applied research into direct numerical control of A-1 reactor temperature

International Nuclear Information System (INIS)

Karpeta, C.; Volf, K.

1974-01-01

Partial results of research efforts aimed at applying modern control theory in the control of the reactor of the A-1 nuclear power station are presented. A mathematical model of the process dynamics was developed. Some parameters of the model were determined using the results of an experimentally performed reactor scram. The optimal stochastic discrete regulator was determined and closed-loop transients were studied. The possibilities of implementing control routines were investigated using the RPP-16 computer. (author)

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2002-01-01

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

Electrosleeve process for in-situ nuclear steam generator repair

International Nuclear Information System (INIS)

Renaud, E.; Brennenstuhl, A.M.; Stewart, D.R.; Gonzalez, F.

2000-01-01

Degradation of steam generator tubing by localized corrosion is a widespread problem in the nuclear industry that can lead to costly forced outages, unit derating, steam generator replacement or even the permanent shutdown of a reactor. In response to the onset of steam generator degradation at Ontario Power Generation's Pickering Nuclear Generating Station (PNGS) Unit 5, and the determined unsuitability of conventional repair methods (mechanically expanded or welded sleeves) for Alloy 400, an alternative repair technology was developed. Electrosleeve is a non-intrusive, low-temperature process that involves the electrodeposition of a nanocrystalline nickel microalloy forming a continuously bonded, structural layer over the internal diameter of the degraded region. This technology is designed to provide a long-term pressure boundary repair, fully restoring the structural integrity of the damaged region to its original state. This paper describes the Electrosleeve process for steam generator tubing repair and the unique properties of the advanced sleeve material. The successful installation of fourteen Electrosleeves that have been in service for more than six years in Alloy 400 tubing at the Pickering-S CANDU unit, and the more recent (Nov. 99) extension of the technology to Alloy 600 by the installation of 57 sleeves in a U.S. pressurized water reactor (PWR) at Callaway, is presented. The Electrosleeve process has been granted a conditional license by the U.S. Nuclear Regulatory Commission (NRC). In Canada, the process of licensing Electrosleeve with the CNSC / TSSA has begun. (author)

Research reactor core conversion guidebook. V.1: Summary

International Nuclear Information System (INIS)

1992-04-01

In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

Reed Reactor Facility final report, September 1, 1995--August 31, 1996

International Nuclear Information System (INIS)

1997-01-01

This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

Reed Reactor Facility final report, September 1, 1995--August 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

International Nuclear Information System (INIS)

Paiva, Rosemeire P.; Salvetti, Tereza C.

2005-01-01

Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

Life extension of the St. Lucie unit 1 reactor vessel

International Nuclear Information System (INIS)

Rowan, G.A.; Sun, J.B.; Mott, S.L.

1991-01-01

In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

Reactor calculations in aid of isotope production at SAFARI-1

International Nuclear Information System (INIS)

Ball, G.

2003-01-01

Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

Shielding assessment of the ETRR-1 Reactor Under power upgrading

Energy Technology Data Exchange (ETDEWEB)

Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1997-12-31

The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

International Nuclear Information System (INIS)

Chambadal, P.; Pascal, M.

1955-01-01

The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr

Development of a training simulator to operators of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carvalho, Ricardo Pinto de

2006-01-01

This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

An overview of the RECH-1 reactor conversion

International Nuclear Information System (INIS)

Klein, J.; Medel, J.; Daie, J.; Torres, H.

2000-01-01

The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

International Nuclear Information System (INIS)

Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

2005-01-01

G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

Verification of the linearity of the IPR-R1 TRIGA reactor power channels

International Nuclear Information System (INIS)

Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

2013-01-01

The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

International Nuclear Information System (INIS)

Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

2011-01-01

JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

An economic analysis of stretch-out for Angra-1 reactor

International Nuclear Information System (INIS)

Sakai, M.

1989-01-01

An application of NUCOST code for calculating nuclear energy cost is presented. Ann optimization of stretch-out for Angra-1 reactor based on international costs of nuclear fuel, operation and maintenance is done. (M.C.K.)

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

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

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

Welding electrode for peripheral welds of A-1 reactor pressure vessel

International Nuclear Information System (INIS)

Lakatos, L.

1975-01-01

The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)

1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Kim, Jae Cheol

2007-02-01

During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

Performance of Canadian commercial nuclear units and heavy water plants

International Nuclear Information System (INIS)

Woodhead, L.W.; Ingolfsrud, L.J.

The operating history of Canadian commercial CANDU type reactors, i.e. Pickering generating station-A, is described. Capacity factors and unit energy costs are analyzed in detail. Equipment performance highlights are given. The performance of the two Canadian heavy water plants is described and five more are under construction or planned. (E.C.B.)

The reactor kinetics code tank: a validation against selected SPERT-1b experiments

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

Energy Technology Data Exchange (ETDEWEB)

Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

2003-07-01

By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

. Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

International Nuclear Information System (INIS)

Yusuf, I; Mati, A. A.

2010-01-01

The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

International Nuclear Information System (INIS)

Matejka, K.; Sklenka, L.

2003-01-01

The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

International Nuclear Information System (INIS)

Auterinen, I.; Salmenhaara, S.E.J. . Author

2004-01-01

The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

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

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

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

Nuclear material control at IEA-R1 nuclear research reactor

International Nuclear Information System (INIS)

1988-01-01

The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

Energy Technology Data Exchange (ETDEWEB)

Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

1999-12-31

Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

Reed Reactor Facility final report, September 1, 1994--August 31, 1995

International Nuclear Information System (INIS)

1997-01-01

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits

Reed Reactor Facility final report, September 1, 1994--August 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

Self-Assembled Colloidal Particle Clusters from In Situ Pickering-Like Emulsion Polymerization via Single Electron Transfer Mechanism.

Science.gov (United States)

Yuan, Jinfeng; Zhao, Weiting; Pan, Mingwang; Zhu, Lei

2016-08-01

A simple route is reported to synthesize colloidal particle clusters (CPCs) from self-assembly of in situ poly(vinylidene fluoride)/poly(styrene-co-tert-butyl acrylate) [PVDF/P(St-co-tBA)] Janus particles through one-pot seeded emulsion single electron transfer radical polymerization. In the in situ Pickering-like emulsion polymerization, the tBA/St/PVDF feed ratio and polymerization temperature are important for the formation of well-defined CPCs. When the tBA/St/PVDF feed ratio is 0.75 g/2.5 g/0.5 g and the reaction temperature is 35 °C, relatively uniform raspberry-like CPCs are obtained. The hydrophobicity of the P(St-co-tBA) domains and the affinity of PVDF to the aqueous environment are considered to be the driving force for the self-assembly of the in situ formed PVDF/P(St-co-tBA) Janus particles. The resultant raspberry-like CPCs with PVDF particles protruding outward may be promising for superhydrophobic smart coatings. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

International Nuclear Information System (INIS)

Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

2010-01-01

In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

1DB, a one-dimensional diffusion code for nuclear reactor analysis

International Nuclear Information System (INIS)

Little, W.W. Jr.

1991-09-01

1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs

Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-08-15

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

Dominant accident sequences in Oconee-1 pressurized water reactor

International Nuclear Information System (INIS)

Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

1985-04-01

A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

RA reactor operation and maintenance in 1989, Part 1

International Nuclear Information System (INIS)

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

1989-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

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

System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

International Nuclear Information System (INIS)

Ellis, R.J.

2000-01-01

The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1988-01-01

Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

Core calculations for the upgrading of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

1998-01-01

The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

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

Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

2005-01-01

The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)

Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

Energy Technology Data Exchange (ETDEWEB)

Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

2014-08-15

IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Operating reactors licensing actions summary. Volume 5, Number 1

International Nuclear Information System (INIS)

1985-03-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Fuel channel design improvements for large CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

International Nuclear Information System (INIS)

Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

2008-01-01

The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

Energy Technology Data Exchange (ETDEWEB)

Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

2008-10-15

The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

Source term determination from subcritical multiplication measurements at Koral-1 reactor

International Nuclear Information System (INIS)

Blazquez, J.B.; Barrado, J.M.

1978-01-01

By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

International Nuclear Information System (INIS)

Axtmann, R.C.; Perkins, H.K.

1977-08-01

Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems

Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

Directory of Open Access Journals (Sweden)

Muhammad Atta

2011-01-01

Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Effects of variable-row-spacing harvesting picker platform scraping plates on cotton fiber quality and quantity

Directory of Open Access Journals (Sweden)

Cíntia Michele de Campos Baraviera

2017-06-01

Full Text Available There have been increasing demands for high-quality cotton fibers that meet the textile industry quality standards. Concurrently, there have been efforts to reduce contaminants during harvesting to reduce harvesting costs. The goal of this research was to evaluate the efficiency of the picker platform with Variable-Row-Spacing (VRS for harvesting cotton in narrow rows, over two harvest seasons in two regions within the state of Mato Grosso, Brazil. In this study, how the presence vs. absence of scraping plates and variations in travel speed was related to quantifiable levels of impurities the harvested fibers was examined. The research was divided into three experiments (Exp. I, II, and III, using cotton varieties FM 975 WS, IMA 5672 B2 RF, and IMA 5675 B2 RF, with row spacing of 0.45 m. The experimental design was randomized blocks, in a 2 ? 3 factorial design, using the presence/absence of the plate and three speeds (0.61, 1.0, and 1.42 m·s-¹, with seven repetitions, totaling 42 experimental plots. The plot size was 108 m² (3.6 ? 30 m. The data were analyzed using the F test in ANOVA and the post-hoc Tukey test (p < 0.05. The results showed that scraping plates increased the number of stems and cones, and reduced the harvest efficiency of cotton planted in narrow rows in the region of Sorriso-MT during the 2013/2014 harvest. For the 2014/2015 harvest, the highest speed and the presence of the scraping plates increased the number of cones in the cotton samples. In the experiment conducted in Serra da Petrovina, the removal of the scraping plates decreased the amount of cones in the harvested cotton.

Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

International Nuclear Information System (INIS)

D'Utra Bitelli, U.

1993-01-01

This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

International Nuclear Information System (INIS)

Hansen, W.

2003-01-01

Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

International Nuclear Information System (INIS)

Balogun, G.I.; Jonah, S.A.; Umar, I.M.

2005-01-01

Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

International Nuclear Information System (INIS)

1995-01-01

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period

Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-31

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)