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.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro`s Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations` safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs.

AECB staff annual assessment of the Pickering A and B Nuclear Generating Stations for the year 1994

International Nuclear Information System (INIS)

1995-06-01

The Pickering Nuclear Generating Station (PNGS) is located on the north shore of Lake Ontario, about 32 km east of downtown Toronto. It consists of two stations, PNGS-A and PNGS-B. Each station contains four reactor units. PNGS-A consists of Units 1 to 4, while PNGS-B consists of Units 5 to 8. Each unit can generate about 540 megawatts of electricity. All eight units are located within a single enclosure. Ontario Hydro's Pickering Nuclear Division has assigned one Station Director with authority over both stations, but each station has its own organization. AECB issue a separate operating licence for each station. This report presents the Atomic Energy Control Board staff assessment of the Pickering stations' safety performance in 1994 and other aspects that they consider to have significant impact on nuclear safety. AECB based their conclusions on their observations, audits, inspections and review of information that Ontario Hydro submits to them as required by the station Operating Licences. 11 tabs., 8 figs

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

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.

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)

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.

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.

Evaluation of organic coatings to reduce air leakage through cracks in the Pickering NGS 'A' reactor building 1

International Nuclear Information System (INIS)

Deans, J.J.; Sato, J.A.; Hampton, J.H.D.; Cullen, R.; Paterson, G.; Chan, P.; Rajagopalan, R.

1994-01-01

Pressure tests conducted in 1992 on the Pickering NGS 'A' Reactor Building 1 showed that the containment leakage rate of the building was close to the licensing limit. The leakage was found to be pressure dependent and was attributed to cracks in the concrete dome. A number of solutions were studied by a task group, and the application of an organic coating to the exterior surface of the dome was identified as the most viable solution under the constraints of schedule and cost. In addition to reducing the air leakage rate, the coating material must be flexible to bridge existing moving cracks, it must have excellent adhesion to the concrete substrate to sustain the design pressure of 41.4 kPa(g) during pressure tests, and it must be durable for an exterior application and service conditions. Five candidate organic coating materials were selected for laboratory testing. As a result of the testing, a single-component elastomeric polyurethane coating was selected to be used on the dome. This paper discusses the selection process, laboratory tests and results, and the application of the polyurethane coating system to the exterior concrete dome surface. However, the main emphasis of the paper is on the laboratory evaluation of the five candidate materials. (author). 2 refs., 3 tabs., 1 fig

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.)

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)

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

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)

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...

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.

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.)

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

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.

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.

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

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)

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)

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

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)

Pickering NGS A reactor building 1 dome refurbishment long-term monitoring of coating

International Nuclear Information System (INIS)

Deans, J.J.; Chan, P.; Gomme, R.

2006-01-01

'Full text:' To reduce air leakage through the dome of Pickering NGS A Reactor Building 1, in August 1993 a portion of the exterior concrete surface was coated with a single component elastomeric polyurethane material. An internal positive pressure test of the building, conducted between November 5 and 7, 1993, found that the air leakage rates were significantly lower in this test than leakage rates which had been measured during a pressure test conducted in 1992. This reduction in leakage was attributed to the successful performance of the coating. The need for a high-performance, elastomeric surface coating was identified for reduction of air leakage levels through the dome of Reactor Building l of Ontario Power Generation's (formerly Ontario Hydro's) Pickering 'A' Nuclear Generating Station near Toronto. A number of candidate coatings were extensively tested to assess the performance characteristics and identify a material that could withstand the elements and perform effectively for around 20 years. Under normal operating conditions, a licensing limit of 2.7% of contained mass/hour is set for permissible containment leakage whilst the operational working target is less than 1%. The facility's engineers determined that any leakages were pressure-dependent, so in an effort to remain well within their working target, they sought a system that would bridge and seal any hairline cracks in the concrete dome and thereby prevent the passage of gas or vapour through the substrate. On the basis of scheduling and cost, they concluded that a high performance coating was most appropriate for the project, and hired Kinectrics (formerly Ontario Hydro Technologies (OHT)) to select, test, assess and arrange for the application to the RB 1 Dome. In all, nearly 70 separate manufacturers were approached by Kinectrics with a view to obtaining recommendations for treatment. The respective performance data of the respondents' products were compared with a set of specific design

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.

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.

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.

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)

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...

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

Pickering Unit 1 chemical cleaning

International Nuclear Information System (INIS)

Smee, J.L.; Fiola, R.J.; Brennenstuhl, K.R.; Zerkee, D.D.; Daniel, C.M.

1995-01-01

The secondary sides of all 12 boilers at Pickering Unit 1 were chemically cleaned in 1994 by the team of Ontario Hydro, B and W International (Cambridge, Ontario) and B and W Nuclear Technologies (Lynchburg, Virginia). A multi-step EPRI/SGOG process was employed in a similar manner to previous clearings at Units 5 and 6 in 1992 and 1993, respectively. A major innovation with the Unit 1 cleaning was the incorporation of a crevice cleaning step, the first time this had been done on Ontario Hydro plants. In addition, six boilers were cleaned in parallel compared to three at a time in previous Pickering cleanings. This significantly reduced cleaning time. A total of 6,770 kg of sludge was removed through direct chemical dissolution. It consisted of 66% iron/nickel oxides and 28% copper metal. A total of 1,600,000 L (420,000 US gallons) of liquid waste was produced. It was processed through the spent solvent treatment facility located at the Bruce Nuclear Power Development site. Visual inspection performed after the cleaning indicated that the crevices between the boiler tubes and the tube support structure were completely clear of deposit and the general condition of the tubing and lattice bars appeared to be in 'as new' condition. (author)

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)

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.)

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

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.

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

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.

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

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.

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.

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

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.

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

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

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.

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.

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.

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.

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)

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...

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).

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

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

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)

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.

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.

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...

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.

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

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)

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

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)

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.

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.

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.

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)

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.

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.

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

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)

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.

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

Fabrication of magnetic hydroxypropyl cellulose-g-poly(acrylic acid) porous spheres via Pickering high internal phase emulsion for removal of Cu(2+) and Cd(2.).

Science.gov (United States)

Zhu, Yongfeng; Zheng, Yian; Zong, Li; Wang, Feng; Wang, Aiqin

2016-09-20

A series of magnetic hydroxypropyl cellulose-g-poly(acrylic acid) porous spheres were prepared via O/W Pickering high internal phase emulsions (HIPEs) integrated precipitation polymerization. The structure and composition of modified Fe3O4 and porous structures were characterized by TEM, XRD, TGA and SEM. The results indicated that the silanized Fe3O4 can influence greatly the pore structure of magnetic porous sphere in addition to non-negligible impacts of the proportion of mixed solvent and co-surfactant. The adsorption experiment demonstrated that the adsorption equilibrium can be reached within 40min and the maximal adsorption capacity was 300.00mg/g for Cd(2+) and 242.72mg/g for Cu(2+), suggesting its fast adsorption kinetics and high adsorption capacity. After five adsorption-desorption cycles, no significant changes in the adsorption capacity were observed, suggesting its excellent reusability. The magnetic porous sphere can be easily separated from the solution and then find its potential as a recyclable material for highly efficient removal of heavy metals. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

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

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.

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

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

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Reactor BR2: Introduction

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-15

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

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.

EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

International Nuclear Information System (INIS)

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

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.

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

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)

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

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.

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

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

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

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

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

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

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

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.

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

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.

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (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.

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

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.

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.

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

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.

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

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.

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)

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

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

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

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

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (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)

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

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.)

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

BR2 reactor neutron beams

International Nuclear Information System (INIS)

Neve de Mevergnies, M.

1977-01-01

The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (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%

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

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

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.

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)

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

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

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

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.

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

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

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

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

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

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%.

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.

TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

International Nuclear Information System (INIS)

Kaczynski, G.M.; Woodruff, R.W.

1985-12-01

TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

Once-through CANDU reactor models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.

1980-11-01

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1988-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

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.)

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

Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

International Nuclear Information System (INIS)

Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

1981-05-01

The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

1998-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

Active species in a large volume N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

2007-01-01

A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

Reactor handbook. 2. rev. ed.

International Nuclear Information System (INIS)

Lederer, B.J.; Wildberg, D.W.

1992-01-01

On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

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

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)

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

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

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

Highlights of the metallurgical behaviour of CANDU pressure tubes

International Nuclear Information System (INIS)

Price, E.G.

1984-10-01

This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

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

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)

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)

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

International Nuclear Information System (INIS)

Townes, B.M.; Hilborn, J.W.

1985-06-01

A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

Operation of the SLOWPOKE-2 reactor in Jamaica

Energy Technology Data Exchange (ETDEWEB)

Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

2001-07-01

Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

Irradiation techniques at BR2 reactor

International Nuclear Information System (INIS)

Hebel, W.

1978-01-01

Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (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.

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

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.

Refurbishment programme for the BR2-reactor

Energy Technology Data Exchange (ETDEWEB)

Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

1992-07-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Refurbishment programme for the BR2-reactor

International Nuclear Information System (INIS)

Koonen, E.

1992-01-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Refurbishing the BR2 materials testing reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Dekeyser, J.; Gubel, P.

1995-01-01

SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

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.

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

International Nuclear Information System (INIS)

1966-01-01

This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

SIRIUS 2: A versatile medium power research reactor

International Nuclear Information System (INIS)

Rousselle, P.

1992-01-01

Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

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.

OTUS - Reactor inventory management system based on ORIGEN2

Energy Technology Data Exchange (ETDEWEB)

Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

1995-10-01

ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

The 2nd reactor core of the NS Otto Hahn

International Nuclear Information System (INIS)

Manthey, H.J.; Kracht, H.

1979-01-01

Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

International Nuclear Information System (INIS)

Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

2010-01-01

The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

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

Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

International Nuclear Information System (INIS)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

1983-02-01

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

International Nuclear Information System (INIS)

Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

2014-01-01

An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

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...

Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Villeger, S; Cousty, S; Ricard, A; Sixou, M

2003-01-01

Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

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)

Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

International Nuclear Information System (INIS)

ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

2017-01-01

The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

International Nuclear Information System (INIS)

Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

2004-01-01

As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

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.

Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

International Nuclear Information System (INIS)

Batista, J.L.; Renke, C.A.C.

1991-11-01

This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

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)

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Radionuclide distribution in TMI-2 reactor building basement liquids and solids

International Nuclear Information System (INIS)

Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

1984-01-01

As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

An experimental investigation of fission product release in SLOWPOKE-2 reactors

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

International Nuclear Information System (INIS)

Hurt, S.S.

1986-11-01

The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

International Nuclear Information System (INIS)

Chatt, A.; Holzbecher, J.

1990-01-01

SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

International Nuclear Information System (INIS)

Greene, S.R.; Spellman, D.J.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

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)

Dynamic simulation of the 2 MWt slowpoke heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1982-04-01

A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

Distribution of energy of impulses of the modernized IBR-2 REACTOR

International Nuclear Information System (INIS)

Tayibov, L.A; Mehtiyeva, R.N.; )

2011-01-01

Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

A conceptual design of LIB fusion reactor: UTLIF(2)

International Nuclear Information System (INIS)

Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

1984-01-01

UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

The BR2 materials testing reactor. Past, ongoing and under-study upgradings

Energy Technology Data Exchange (ETDEWEB)

Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

1990-05-01

The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

Directory of Open Access Journals (Sweden)

Yasmine Abdel-Maksoud

2016-09-01

Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

Planned Scientific programs around the Triga Mark 2 Reactor

International Nuclear Information System (INIS)

Majah, M Ibn.

2007-01-01

Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

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)

Power noise spectrum classification in the problem of the IBR-2 reactor

International Nuclear Information System (INIS)

Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

1988-01-01

The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

Research reactor FR2 - 20 years chemical and radiochemical measurements

International Nuclear Information System (INIS)

Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

1986-09-01

The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

Venting krypton-85 from the Three Mile Island Unit 2 reactor building

International Nuclear Information System (INIS)

Burton, H.M.

1981-01-01

To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

Research on economics and CO2 emission of magnetic and inertial fusion reactors

International Nuclear Information System (INIS)

Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

2011-01-01

An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

Development of a TiO2-coated optical fiber reactor for water decontamination

International Nuclear Information System (INIS)

Danion, A.

2004-09-01

The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

TMI-2 reactor-vessel head removal and damaged-core-removal planning

International Nuclear Information System (INIS)

Logan, J.A.; Hultman, C.W.; Lewis, T.J.

1982-01-01

A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

International Nuclear Information System (INIS)

Ipek, O.

1997-01-01

The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

NARCIS (Netherlands)

Gallucci, F.; Basile, A.

2007-01-01

In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

Cronos 2: a neutronic simulation software for reactor core calculations

International Nuclear Information System (INIS)

Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

1999-01-01

The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

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.)

Set of rules SOR 2 reactor site criteria

International Nuclear Information System (INIS)

1976-06-01

The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-01-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

International Nuclear Information System (INIS)

Richard P. Wells

2007-01-01

This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-11-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

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.

Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

International Nuclear Information System (INIS)

Abou Yehia, H.; Bars, G.; Tran Dai

2001-01-01

After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Selective separation of lambdacyhalothrin by porous/magnetic molecularly imprinted polymers prepared by Pickering emulsion polymerization.

Science.gov (United States)

Hang, Hui; Li, Chunxiang; Pan, Jianming; Li, Linzi; Dai, Jiangdong; Dai, Xiaohui; Yu, Ping; Feng, Yonghai

2013-10-01

Porous/magnetic molecularly imprinted polymers (PM-MIPs) were prepared by Pickering emulsion polymerization. The reaction was carried out in an oil/water emulsion using magnetic halloysite nanotubes as the stabilizer instead of a toxic surfactant. In the oil phase, the imprinting process was conducted by radical polymerization of functional and cross-linked monomers, and porogen chloroform generated steam under the high reaction temperature, which resulted in some pores decorated with easily accessible molecular binding sites within the as-made PM-MIPs. The characterization demonstrated that the PM-MIPs were porous and magnetic inorganic-polymer composite microparticles with magnetic sensitivity (M(s) = 0.7448 emu/g), thermal stability (below 473 K) and magnetic stability (over the pH range of 2.0-8.0). The PM-MIPs were used as a sorbent for the selective binding of lambdacyhalothrin (LC) and rapidly separated under an external magnetic field. The Freundlich isotherm model gave a good fit to the experimental data. The adsorption kinetics of the PM-MIPs was well described by pseudo-second-order kinetics, indicating that the chemical process could be the rate-limiting step in the adsorption of LC. The selective recognition experiments exhibited the outstanding selective adsorption effect of the PM-MIPs for target LC. Moreover, the PM-MIPs regeneration without significant loss in adsorption capacity was demonstrated by at least four repeated cycles. © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

1989-01-01

A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

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.

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Rapid data acquisition from the safety system of the FRJ-2 reactor

International Nuclear Information System (INIS)

Inhoven, H.

1980-06-01

The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

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

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

Estimation of power feedback parameters of pulse reactor IBR-2M on transients

International Nuclear Information System (INIS)

Pepyolyshev, Yu.N.; Popov, A.K.

2013-01-01

Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

International Nuclear Information System (INIS)

Stojic, M.; Pavicevic, M.

1964-01-01

This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

PSA Level 2 activities for RBMK reactors

International Nuclear Information System (INIS)

Gubler, R.

1998-01-01

Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

Studsvik's R2 reactor - Review of activities

Energy Technology Data Exchange (ETDEWEB)

Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

1993-07-01

A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

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.

Annual report/1979

International Nuclear Information System (INIS)

1980-04-01

Primary energy demand in Ontario in 1979 was up by 2.9 percent, compared to 2.7 percent in the previous year. Revised forecasts issued in January 1980 indicate Ontario's need for electricity is expected to grow by an average of 3.4 percent annually to the year 2000. Nuclear generation provided 29 percent of the total energy made available by Hydro, and Hydro's eight reactors at Pickering and Bruce continued to rank in the top 36 - four in the top 10 - when compared to the permance of 104 of the world's largest reactors. The provinical legislature's Select Committee on Hydro Affairs examined the safety of the CANDU system and concluded that is is 'acceptably safe'. Faced with reduced forecasts of electrical demands to the year 2000 the Board of Directors decided to stretch out the construction schedule of the Darlington station, to halt construction of the second half of the Bruce Heavy Water Plant D, and to complete but mothball the first half. Construction of Bruce Heavy Water Plant B was completed early in 1979. The A plant produced 599.8 megagrams of reactor-grade heavy water. A control room simulator for Bruce A nuclear generating station was ordered. Agreement was reached on rebuilding faulty boilers at Pickering B. A total of 757.6 megagrams of uranium was used to produce electrical energy and steam. Ontario Hydro continued involvement in uranium exploration. Studies on radioactive waste disposal are being carried out, with emphasis on interim storage and transportation. (LL)

Decommissioning of reactor facilities (2). Required technology

International Nuclear Information System (INIS)

Yanagihara, Satoshi

2014-01-01

Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1983-08-15

To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

Production of Sn-117m in the BR2 high-flux reactor.

Science.gov (United States)

Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

2009-01-01

The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

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

Comparative study between fluidized bed and fixed bed reactors in methane reforming with CO2 and O2 to produce syngas

International Nuclear Information System (INIS)

Jing Qiangshan; Lou Hui; Mo Liuye; Zheng Xiaoming

2006-01-01

Reforming of methane with carbon dioxide and oxygen was investigated over Ni/MgO-SiO 2 catalysts using fixed bed and fluidized bed reactors. The conversions of CH 4 and CO 2 in a fluidized bed reactor were close to thermodynamic equilibrium. The activity and stability of the catalyst in the fixed bed reactor were lower than that in the fluidized bed reactor due to carbon deposition and nickel sintering. TGA and TEM techniques were used to characterize the spent catalysts. The results showed that a lot of whisker carbon was found on the catalyst in the rear of the fixed bed reactor, and no deposited carbon was observed on the catalysts in the fluidized bed reactor after reaction. It is suggested that this phenomenon is related to a permanent circulation of catalyst particles between the oxygen rich and oxygen free zones. That is, fluidization of the catalysts in the fluidized bed reactor favors inhibiting deposited carbon and thermal uniformity in the reactor

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

Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

International Nuclear Information System (INIS)

Burley, E.L.

1982-07-01

A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

Characterization of fuel distributions in the Three-Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

1984-04-01

The resolution of technical issues generated by the accident at Three-Mile Island Unit 2 (TMI-2) will inevitably be of long range benefit. Determination of the fuel debris dispersal in the TMI-2 reactor system represents a major technical issue. In reactor recovery operations, such as for the safe handling and final disposal of TMI-2 waste, quantitative fuel assessments are being conducted throughout the reactor core and primary coolant system

Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

International Nuclear Information System (INIS)

Bissani, M; O'Kelly, D S

2006-01-01

A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

Energy Technology Data Exchange (ETDEWEB)

Bissani, M; O' Kelly, D S

2006-05-08

A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

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)

analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

International Nuclear Information System (INIS)

Elnokity, O.E.M.

2006-01-01

this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

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.)

General outline of the operation and utilization of the BR2 reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

1978-01-01

The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

Benchmark testing of CENDL-2 for U-fuel thermal reactors

International Nuclear Information System (INIS)

Zhang Baocheng; Liu Guisheng; Liu Ping

1995-01-01

Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

The 5th surveillance testing for Kori unit 2 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-03-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

Digital, remote control system for a 2-MW research reactor

International Nuclear Information System (INIS)

Battle, R.E.; Corbett, G.K.

1988-01-01

A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

A nodal Grean's function method of reactor core fuel management code, NGCFM2D

International Nuclear Information System (INIS)

Li Dongsheng; Yao Dong.

1987-01-01

This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

RHTF 2, a 1200 MWe high temperature reactor

International Nuclear Information System (INIS)

Brisbois, Jacques

1978-01-01

After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

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)

Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

International Nuclear Information System (INIS)

Guelol Oezdere, Oya

2003-01-01

Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

International Nuclear Information System (INIS)

Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

2016-01-01

Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

International Nuclear Information System (INIS)

Collins, N.; Lafreniere, P.

1991-01-01

In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

Maximum credible accident analysis for TR-2 reactor conceptual design

International Nuclear Information System (INIS)

Manopulo, E.

1981-01-01

A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

Energy Technology Data Exchange (ETDEWEB)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2016-03-11

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

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

Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

International Nuclear Information System (INIS)

2015-01-01

This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

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)

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

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

1987-01-01

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

A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

2014-01-01

A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

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

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

International Nuclear Information System (INIS)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

2013-01-01

This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

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.

Severe accident analysis for level 2 PSA of SMART reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

2010-12-15

The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

International Nuclear Information System (INIS)

Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

1999-01-01

Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

Energy Technology Data Exchange (ETDEWEB)

Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

2017-01-15

Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

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

Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant.

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)

Problems of nuclear reactor safety. Vol. 2

International Nuclear Information System (INIS)

Goncharov, L.A.

1995-01-01

Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

International Nuclear Information System (INIS)

Aziz, M.; El Messiry, A.M.

2000-01-01

The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

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

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

2012-01-01

CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

Science.gov (United States)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Utilization of the SLOWPOKE-2 research reactor

International Nuclear Information System (INIS)

Lalor, G.C.

2001-01-01

SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (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.)

Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

International Nuclear Information System (INIS)

Markovic, H.

1962-07-01

The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Souza Mendes, J.E. de

1987-01-01

Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

International Nuclear Information System (INIS)

Barelli, L.; Bidini, G.; Gallorini, F.

2016-01-01

Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

Directory of Open Access Journals (Sweden)

Abdul Wahid

2015-06-01

Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

Operating reactors licensing actions summary. Volume 5, No. 2

International Nuclear Information System (INIS)

1985-04-01

The Operating Reactors Licensing Actions Summary 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

Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

Energy Technology Data Exchange (ETDEWEB)

Nilsson, R; Aalto, E

1964-09-15

Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

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.)

Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

Directory of Open Access Journals (Sweden)

Sylvestre Bachollet

2013-10-01

Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

International Nuclear Information System (INIS)

Abd El-Rahman, M.F.

2004-01-01

Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

Thermal design of heat-exchangeable reactors using a dry-sorbent CO2 capture multi-step process

International Nuclear Information System (INIS)

Moon, Hokyu; Yoo, Hoanju; Seo, Hwimin; Park, Yong-Ki; Cho, Hyung Hee

2015-01-01

The present study proposes a multi-stage CO 2 capture process that incorporates heat-exchangeable fluidized-bed reactors. For continuous multi-stage heat exchange, three dry regenerable sorbents: K 2 CO 3 , MgO, and CaO, were used to create a three-stage temperature-dependent reaction chain for CO 2 capture, corresponding to low (50–150 °C), middle (350–650 °C), and high (750–900 °C) temperature stages, respectively. Heat from carbonation in the high and middle temperature stages was used for regeneration for the middle and low temperature stages. The feasibility of this process is depending on the heat-transfer performance of the heat-exchangeable fluidized bed reactors as the focus of this study. The three-stage CO 2 capture process for a 60 Nm 3 /h CO 2 flow rate required a reactor area of 0.129 and 0.130 m 2 for heat exchange between the mid-temperature carbonation and low-temperature regeneration stages and between the high-temperature carbonation and mid-temperature regeneration stages, respectively. The reactor diameter was selected to provide dense fluidization conditions for each bed with respect to the desired flow rate. The flow characteristics and energy balance of the reactors were confirmed using computational fluid dynamics and thermodynamic analysis, respectively. - Highlights: • CO 2 capture process is proposed using a multi-stage process. • Reactor design is conducted considering heat exchangeable scheme. • Reactor surface is designed by heat transfer characteristics of fluidized bed

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Core design calculations of IRIS reactor using modified CORD-2 code package

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

2002-01-01

Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(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

BR2 reactor: medical and industrial applications

International Nuclear Information System (INIS)

Ponsard, B.

2005-01-01

The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

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.

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

A regression approach for Zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

International Nuclear Information System (INIS)

Elliot, A.J.; Stuart, C.R.

2008-06-01

An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

2014-01-01

Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

Science.gov (United States)

Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

2014-01-01

The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

Direct In Situ Quantification of HO2 from a Flow Reactor.

Science.gov (United States)

Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

2013-03-21

The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics.

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Status and perspective of development of cold moderators at the IBR-2 reactor

International Nuclear Information System (INIS)

Kulikov, S; Shabalin, E

2012-01-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Status and perspective of development of cold moderators at the IBR-2 reactor

Science.gov (United States)

Kulikov, S.; Shabalin, E.

2012-03-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

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.

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

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

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

Further study on parameterization of reactor NAA: Pt. 2

International Nuclear Information System (INIS)

Tian Weizhi; Zhang Shuxin

1989-01-01

In the last paper, Ik 0 method was proposed for fission interference corrections. Another important kind of interferences in reator NAA is due to threshold reaction induced by reactor fast neutrons. In view of the increasing importance of this kind of interferences, and difficulties encountered in using the relative comparison method, a parameterized method has been introduced. Typical channels in heavy water reflector and No.2 horizontal channel of Heavy Water Research Reactor in the Insitute of Atomic Energy have been shown to have fast neutron energy distributions (E>4 MeV) close to primary fission neutron spectrum, by using multi-threshold detectors. On this basis, Ti foil is used as an 'instant fast neutron flux monitor' in parameterized corrections for threshold reaction interferences in the long irradiations. Constant values of φ f /φ s = 0.70 ± 0.02% have been obtained for No.2 rabbit channel. This value can be directly used for threshold reaction inference correction in the short irradiations

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

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

Integral Inherently Safe Light Water Reactor (I2S-LWR)

International Nuclear Information System (INIS)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

2017-01-01

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

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

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

Conducting a nuclear heart transplant

International Nuclear Information System (INIS)

Mark, Brian; Trzeciakowski, Jan

1992-01-01

The retubing of Pickering 'A' is discussed largely from an organizational point of view, with the emphasis on project management, teamwork, and safety. In 1983, work was started on Units 1 and 2, but it was later found better to retube one reactor at a time. Two robotic arms originally built for the purpose were found to be unsuitable, and it was necessary to adopt a semi-manual approach. Through experience and teamwork, the time taken to retube was lowered from five years for Unit 2 to 26 months for Unit 3, and 18 months for Unit 4

TiO2-photocatalyzed As(III) oxidation in a fixed-bed, flow-through reactor.

Science.gov (United States)

Ferguson, Megan A; Hering, Janet G

2006-07-01

Compliance with the U.S. drinking water standard for arsenic (As) of 10 microg L(-1) is required in January 2006. This will necessitate implementation of treatment technologies for As removal by thousands of water suppliers. Although a variety of such technologies is available, most require preoxidation of As(III) to As(V) for efficient performance. Previous batch studies with illuminated TiO2 slurries have demonstrated that TiO2-photocatalyzed AS(III) oxidation occurs rapidly. This study examined reaction efficiency in a flow-through, fixed-bed reactor that provides a better model for treatment in practice. Glass beads were coated with mixed P25/sol gel TiO2 and employed in an upflow reactor irradiated from above. The reactor residence time, influent As(III) concentration, number of TiO2 coatings on the beads, solution matrix, and light source were varied to characterize this reaction and determine its feasibility for water treatment. Repeated usage of the same beads in multiple experiments or extended use was found to affect effluent As(V) concentrations but not the steady-state effluent As(III) concentration, which suggests that As(III) oxidation at the TiO2 surface undergoes dynamic sorption equilibration. Catalyst poisoning was not observed either from As(V) or from competitively adsorbing anions, although the higher steady-state effluent As(III) concentrations in synthetic groundwater compared to 5 mM NaNO3 indicated that competitive sorbates in the matrix partially hinder the reaction. A reactive transport model with rate constants proportional to incident light at each bead layer fit the experimental data well despite simplifying assumptions. TiO2-photocatalyzed oxidation of As(III) was also effective under natural sunlight. Limitations to the efficiency of As(III) oxidation in the fixed-bed reactor were attributable to constraints of the reactor geometry, which could be overcome by improved design. The fixed-bed TiO2 reactor offers an environmentally

Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

International Nuclear Information System (INIS)

Skogen, F.B.; Stout, R.B.

1977-01-01

Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

Characterization of fuel distribution in the Three Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

1984-01-01

Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. For fuel distribution characterization in the core region, results from SSTR neutron dosimetry exposures in the TMI-2 reactor cavity are presented. These SSTR results are consistent with the presence of a significant amount of fuel debris, equivalent to several fuel assemblies or more, lying at the bottom of the reactor vessel. (Auth.)

Estimation of power feedback parameters of the IBR-2M reactor by square wave reactivity

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Popov, A.K.; Sumkhuu, D.

2016-01-01

Parameters of the IBR-2M reactor power feedback (PFB) are estimated based on the analysis of power transients caused by deliberate square wave reactivity when the pulsed reactor operates in the self-regulation mode. The PFB of the IBR-2M is described by three linear first-order differential equations. Two components of the PFB are responsible for the negative feedback and one, for the positive. The overall feedback is negative, i.e., it has a stabilizing effect for the operation of the reactor. The slowest negative component of the PFB is probably caused by heating of the fuel. Periodically repeated in the process of exploitation, estimation of the PFB parameters is one of the methods to ensure safety operation of the reactor. [ru

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Hatch, A.J.

1975-01-01

1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

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

Report on the operation in 1973 of the FR 2 research reactor

International Nuclear Information System (INIS)

Moeller, I.; Steiger, W.

1975-04-01

Also in 1973, the heavy-water moderated research and testing reactor FR 2 was operated to schedule at 44 MW nominal power. Again, the availability of the plant was slightly improved. Experimental utilization through instrumented irradiation capsules strongly increased as compared to the previous year. Up to 16 capsule test rigs at a time were inserted in the reactor. As to the beam tube experiments, up to 13 experiments covering a total of 18 test rigs were conducted simultaneously at the 12 reasonably usable beam holes. At the beginning of the year all of the positions available were occupied by 5 loop experiments. Isotope production reached its highest value with a total of 2,372 irradiated capsules (1.3% more than the year before). Some remarkable figures characterized the year 1973: On August 16, 1973 ten years of full power operation at a nominal power of 12 and 44 MW, respectively, had been reached. On July 24, 1973 the 50,000th isotope irradiation was performed in the reactor and on December 26, 1973 a total energy release of 100,000 MWd was recorded. Moreover, the 125,000th visitor of the reactor was welcomed on December 6, 1973. (orig./UA) [de

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

International Nuclear Information System (INIS)

Doerge, D.H.; Miller, R.L.; Scotti, K.S.

1986-05-01

This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

Energy Technology Data Exchange (ETDEWEB)

Dougherty, D.; Adams, J. W.

1983-08-01

Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

International Nuclear Information System (INIS)

Dougherty, D.; Adams, J.W.

1983-08-01

Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

International Nuclear Information System (INIS)

Finch, D.R.

1965-01-01

1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

G2 and G3 reactors design; Description des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

International Nuclear Information System (INIS)

Alvares, N.J.; Beason, D.G.; Eidem, G.R.

1982-06-01

About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

A review of the probabilistic safety assessment application to the TR-2 research reactor

International Nuclear Information System (INIS)

Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

2001-01-01

A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.

1985-04-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

1986-01-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

Degradation of gas-phase trichloroethylene over thin-film TiO2 photocatalyst in multi-modules reactor

International Nuclear Information System (INIS)

Kim, Sang Bum; Lee, Jun Yub; Kim, Gyung Soo; Hong, Sung Chang

2009-01-01

The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO 2 . A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

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

Science.gov (United States)

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

2015-05-04

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%. Copyright © 2015 Elsevier B.V. All rights reserved.

Some novel on-power refuelling features of CANDU stations

International Nuclear Information System (INIS)

Erwin, D.; Pendlebury, B.; Watson, J.F.; Welch, A.C.

1976-01-01

Part A of the paper describes the reasons for, and advantages resulting from, the use of flow assisted refuelling in the CANDU type nuclear reactors at the Pickering Generating Station. A separate fuel handling system is used for each reactor unit, as distinct from the system employed at the Bruce Generating station, where the fuel handling system is shared among several units. Part B of the paper describes some of the advantages of the shared concept with particular emphasis on the availability of the fuel handling system. (author)

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

A gas-phase reactor powered by solar energy and ethanol for H2 production

International Nuclear Information System (INIS)

Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

2014-01-01

In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Respiratory health risks and exposure to particulate matter (PM 2.5 ...

African Journals Online (AJOL)

A validated questionnaire for respiratory health was administered to 102 waste pickers and exposure to environmental and personal PM2.5 was evaluated. There was a relatively high prevalence of chronic cough and wheeze amongst all participants (57% and 51% respectively). Males reported a higher frequency of cough ...

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

1985-01-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

Directory of Open Access Journals (Sweden)

Hwanyeal Yu

2015-12-01

Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

Hétérogenéité quantitative et qualitative de la matière organique dans les argiles du Kimmeridgien du val de Pickering (Yorkshire, UK). Cadre sédimentologique et stratigraphique Quantitative and Qualitative Heterogeneity of the Organic Matter Within the Kimmeridge Clay of the Vale of Pickering (Yorkshire, Uk). Sedimentological and Stratigraphical Framework

OpenAIRE

Penn I. E.; Herbin J. P.; Muller C.; Geyssant J. R.; Melieres F.

2006-01-01

Les argiles du Kimméridgien contribuent à l'alimentation de nombreux gisements en mer du Nord. L'étude de quatre puits localisés sur une coupe Est-Ouest dans le Val de Pickering (Bassin de Cleveland) permet d'illustrer dans cette formation l'hétérogénéité de distribution verticale et latérale de la matière organique en terme de quantité et de qualité. La coupe étudiée (35 km de long, 200 m de haut) représente environ 6,5 millions d'années d'histoire géologique. Les analyses minéralogiques mon...

Analysis of key hardware factors and countermeasure for restricting 49-2 swimming pool reactor lifetime

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Yang Xiao; Wang Yiwei; Wang Zhanwen

2013-01-01

Safe operation is the most important factor to determine the lifetime of aged 49-2 swimming pool reactor. In this paper, the hardware factors of lifetime were analyzed, such as the pool concrete aging, corrosion of aluminum container and primary coolant system, and graphite swelling etc., and then the corresponding measures such as surveillance, prevention and maintenance were purposed. The results show that 49-2 swimming pool reactor can continue to operate safely due to that container is safe under 8 degree earthquake, the reactor is safe on flood level of once per millennium, adding dam break, and the ageing condition of primary coolant system and container is acceptable. (authors)

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

International Nuclear Information System (INIS)

Collins, N.; Lafreniere, P.

1994-01-01

In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

International Nuclear Information System (INIS)

Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

2015-01-01

A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

International Nuclear Information System (INIS)

Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

1980-01-01

1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

Energy Technology Data Exchange (ETDEWEB)

Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

2013-07-01

In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

Proceedings of the international topical meeting on advanced reactors safety: Volume 2

International Nuclear Information System (INIS)

1997-01-01

In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01