Tuning Amphiphilicity of Particles for Controllable Pickering Emulsion

Directory of Open Access Journals (Sweden)

Zhen Wang

2016-11-01

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

Pickering NGS end of commercial operations

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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

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.

Ontario Hydro Pickering Generating Station fuel handling system performance

International Nuclear Information System (INIS)

Underhill, H.J.

1986-01-01

The report briefly describes the Pickering Nuclear Generating Station (PNGS) on-power fuel handling system and refuelling cycle. Lifetime performance parameters of the fuelling system are presented, including station incapability charged to the fuel handling system, cost of operating and maintenance, dose expenditure, events causing system unavailability, maintenance and refuelling strategy. It is concluded that the 'CANDU' on-power fuelling system, by consistently contributing less than 1% to the PNGS incapability, has been credited with a 6 to 20% increase in reactor capacity factor, compared to off-power fuelling schemes. (author)

Toxocariasis in waste pickers: a case control seroprevalence study.

Directory of Open Access Journals (Sweden)

Cosme Alvarado-Esquivel

Full Text Available BACKGROUND: The epidemiology of Toxocara infection in humans in Mexico has been poorly explored. There is a lack of information about Toxocara infection in waste pickers. AIMS: Determine the seroepidemiology of Toxocara infection in waste pickers. METHODS: Through a case control study design, the presence of anti-Toxocara IgG antibodies was determined in 90 waste pickers and 90 age- and gender-matched controls using an enzyme-linked immunoassay. Associations of Toxocara exposure with socio-demographic, work, clinical, and behavioral data of the waste pickers were also evaluated. RESULTS: The seroprevalence of anti-Toxocara IgG antibodies was significantly higher in waste pickers (12/90: 13% than in control subjects (1/90: 1% (OR = 14; 95% CI: 2-288. The seroprevalence was not influenced by socio-demographic or work characteristics. In contrast, increased seroprevalence was found in waste pickers suffering from gastritis, and reflex and visual impairments. Multivariate analysis showed that Toxocara exposure was associated with a low frequency of eating out of home (OR = 26; 95% CI: 2-363 and negatively associated with consumption of chicken meat (OR = 0.03; 95% CI: 0.003-0.59. Other behavioral characteristics such as animal contacts or exposure to soil were not associated with Toxocara seropositivity. CONCLUSIONS: 1 Waste pickers are a risk group for Toxocara infection. 2 Toxocara is impacting the health of waste pickers. This is the first report of Toxocara exposure in waste pickers and of associations of gastritis and reflex impairment with Toxocara seropositivity. Results warrant for further research.

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

International Nuclear Information System (INIS)

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

1978-12-01

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

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.

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

Commissioning quality assurance at Pickering NGS

International Nuclear Information System (INIS)

Wieckowski, J.T.

1983-05-01

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

Pickering tool management system

International Nuclear Information System (INIS)

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

1997-01-01

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

Seismic analysis of mechanical systems at Pickering NGS

International Nuclear Information System (INIS)

Ghobarah, A.

1995-11-01

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

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

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

1997-01-01

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

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

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

1997-01-01

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

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 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.)

Data-driven warehouse optimization : Deploying skills of order pickers

NARCIS (Netherlands)

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

2015-01-01

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

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

DEFF Research Database (Denmark)

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

2010-01-01

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

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 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

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

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

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.

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

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.

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.

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.

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.

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

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.

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.

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.

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

Directory of Open Access Journals (Sweden)

Kaushal Rege

2010-02-01

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

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

DEFF Research Database (Denmark)

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

2010-01-01

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

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)

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)

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)

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.

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.

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

Science.gov (United States)

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

2018-02-01

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

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

International Nuclear Information System (INIS)

Serdula, K.J.

1988-01-01

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

Pickering emulsions for skin decontamination.

Science.gov (United States)

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

2016-08-01

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

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

Directory of Open Access Journals (Sweden)

Hagspihl, Robert

2014-10-01

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

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

Science.gov (United States)

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

2018-02-01

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

Bruce used fuel dry storage project evolution from Pickering to Bruce

International Nuclear Information System (INIS)

Young, R.E.

1996-01-01

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

Bruce used fuel dry storage project evolution from Pickering to Bruce

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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)

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.

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

Seismic assessment of the Pickering pressure relief duct

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

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

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

Science.gov (United States)

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

2017-09-11

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

Fabrication and evaluation of chitosan/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} upconversion nanoparticles composite beads based on the gelling of Pickering emulsion droplets

Energy Technology Data Exchange (ETDEWEB)

Yan, Huiqiong; Chen, Xiuqiong; Shi, Jia; Shi, Zaifeng; Sun, Wei; Lin, Qiang, E-mail: linqianggroup@163.com; Wang, Xianghui; Dai, Zihao

2017-02-01

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{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs, synthesized by a hydrothermal method, were coated with F127 and then assembled with chitosan to fabricate the chitosan/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} composite beads (CS/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 3+} CBs) by Pickering emulsion system. The characterization results revealed that the as-synthesized NaYF{sub 4}:Yb{sup 3+}/Tm{sup 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{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs were successfully conjugated on the surface of chitosan beads by the gelling of emulsion droplets. The resultant CS/NaYF{sub 4}:Yb{sup 3+}/Tm{sup 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{sub 4}:Yb{sup 3+}/Tm{sup 3+} UCNPs were coated by F127 to improve aqueous dispersibility. • NaYF{sub 4}:Yb{sup 3+}/Tm{sup 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.

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.

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.

Fuel recycling and 4. generation reactors

International Nuclear Information System (INIS)

Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

2012-01-01

The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

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

International Nuclear Information System (INIS)

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

1980-02-01

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

Condition based maintenance pilot projects at Pickering ND

International Nuclear Information System (INIS)

Zemdegs, R.T.

1995-01-01

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

Picker versus stripper harvesters on the High Plains of Texas

Science.gov (United States)

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

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

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

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)

Pipe support program at Pickering

International Nuclear Information System (INIS)

Sahazizian, L.A.; Jazic, Z.

1997-01-01

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

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Science.gov (United States)

Tiberi, Lara; Spallarossa, Daniele; Costa, Giovanni

2013-04-01

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

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

OpenAIRE

Hebalkar, Dr. Rashmi; Sharma, Meena Sunildutt

2013-01-01

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

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

International Nuclear Information System (INIS)

Johnson, K.C.; Rouleau, J.

1991-10-01

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

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

OpenAIRE

van Zeeland, Angelique J.W.M.

2014-01-01

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

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.

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)

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

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

Science.gov (United States)

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

2017-02-01

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 20nm 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. Copyright © 2016 Elsevier B.V. All rights reserved.

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

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

Directory of Open Access Journals (Sweden)

Robert T. Woodward

2016-09-01

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

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

Science.gov (United States)

Hartmann, Chris

2018-01-01

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

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)

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

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)

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 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

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

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.

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.

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.

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

2011-01-01

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

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)

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

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

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.

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(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

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

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)

Chernobyl: recovery operations and the entombment of Reactor 4

International Nuclear Information System (INIS)

Dalziel, S.P.C.

1988-01-01

The immediate actions taken following the accident at the Chernobyl-number 4 reactor in April 1986 are described. These included actions to put out the fires, initial medical aid and the dropping of sand, lead, dolomite and boron onto the reactor from helicopters. Following this the chamber below the damaged reactor core was filled with concrete to prevent any further explosions or meltdown. The reactor was subsequently entombed in steel and concrete. The evacuation of the surrounding area is also mentioned. (U.K.)

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).

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.

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.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

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)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, A. dos; Nascimento, J.A. do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

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

Directory of Open Access Journals (Sweden)

Tim Schadla-Hall

2000-10-01

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

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)

Nuclear energy. The innovations of the N4 reactor

International Nuclear Information System (INIS)

Anon.

1998-01-01

The coupling to the electric network of the two first units of N4 type reactors, on the site of Chooz in the Ardennes, marks the third great step of the French nuclear programme of PWR type reactors, after the realization of 34 units of 900 MWe and 20 units of 1300 M We. The nuclear boiler N4, realizes a new evolution in power, in performances and in reliability. (N.C.)

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.

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

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.

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.

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, S.; Ponsard, B.; Koonen, E.

2007-01-01

The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

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.

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.

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.

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.

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.

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)

A basic design of SR4 instrumentation and control system for research reactor

International Nuclear Information System (INIS)

Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

2010-01-01

An SR4 instrumentation and control systems of research reactor is the equipment of nuclear research reactors as power protection devices and control systems. The equipment is to monitor safety parameters and process parameters in the state of reactor shut down, start-up, and in operation at fixed power. In the engineering of Instrumentation and control systems SR4 research reactor, its basic design consists of technical specifications of the reactor protection system devices, technical specifications of the reactor power control system devices, technical specifications information system devices, and systems process termination cabling as a support system. This basic design is used as the basis for the preparation of detailed design and subsequent engineering development of instrumentation systems and control system integrated. (author)

Reactor power cutback system test experience at YGN 4

International Nuclear Information System (INIS)

Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

1995-01-01

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

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

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

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)

Application of WIMSD-4 for ''MARIA'' reactor lattice calculations

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

1993-12-01

A general description of the WIMSD-4 lattice code is given with the emphasis on available geometrical models. The difficulties encountered while modelling reactor lattices with the tubular type fuel elements are explained. Then the analysis of code options allowing to overcome these difficulties is carried out. Eventually, recommendations of options and input parameters for calculations of MARIA reactor lattice with satisfactory accuracy are given. During the work a set of modifications had to be introduced leading to a new code version called WIMS-S. Another version, under the name WIMS-T has been developed to allow for burnup calculations of the MARIA reactor lattice with improved resonance approach. (author). 14 refs, 6 figs, 10 tabs

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.)

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)

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.)

On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

International Nuclear Information System (INIS)

Platonov, P.A.

2007-01-01

Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

Biodegradation of Jet Fuel-4 (JP-4) in Sequencing Batch Reactors

Science.gov (United States)

1992-06-01

nalw~eo %CUMENTATION PAGE__ _ _ _ _ _ _ _ _O 74S Ab -A258 020 L AW POi~W6 DATI .~ TYP AIMqm ,-& 0 U. glbs A~ I ma"&LFUN Mu BIODEGRADATION OF JET FUEL...Specific Objectives of This Proposal Are: 1. To assess the ability of C. resinae , P. chrysosporium and selected bacterial consortia to degrade individual...chemical components of JP-4. 2. To develop a sequencing batch reactor that utilizes C. resinae to degrade chemical components of JP-4 in contaminated

PNGS-A unit 4 composite graphite valve packing program

International Nuclear Information System (INIS)

Meaney, F.

1992-01-01

During the outage of Pickering 4 for retubing and rehabilitation, a composite graphite packing program was completed on 100 conventional valves and 50 nuclear valves. This paper describes component refurbishment, related maintenance, and manpower requirements. 2 figs

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

Experiment operations plan for the MT-4 experiment in the NRU reactor

International Nuclear Information System (INIS)

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

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.

Fast breeder reactors--lecture 4

International Nuclear Information System (INIS)

Marshall, W.; Davies, L.M.

1986-01-01

This paper discusses the economics of fast breeder reactors. An algebraic background is presented which represents the various views expressed by different nations regarding the cost of fast breeder reactors and their associated fuel cycle services, the timescale by which they might be available, and the simultaneous variations in the price of uranium. Actual presentations made by individual countries in recent discussions serve to verify the general nature of this present discussion. It is assumed that if nuclear power is to make a long term contribution to the needs of the world, the introduction of fast breeder reactors is both essential and necessary

In-reactor oxidation of zircaloy-4 under low water vapor pressures

Science.gov (United States)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

International Nuclear Information System (INIS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

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...

Operating reactors licensing actions summary. Vol.4, No. 4

International Nuclear Information System (INIS)

1984-06-01

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

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

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.

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

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 4th 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 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

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

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

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

Questions about the reactor accident with Chernobyl-4

International Nuclear Information System (INIS)

Heijboer, R.J.

1986-01-01

The author presents an inventory of existing information about the Chernobyl-4 accident. Several possible scenarios are described and a comparison is drawn with the Three Mile Island-2 accident. The author concludes that the event is connected to an inherent instability of the RBMK-1000 reactor type. (G.J.P.)

Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

International Nuclear Information System (INIS)

Orso, J; Marenzana, A

2012-01-01

Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

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

Science.gov (United States)

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

2017-07-01

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

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

Upper limit to the deuterium abundance and a measurement of the pickering-β line in the low excitation planetary IC 418

International Nuclear Information System (INIS)

Le Vaux, H.A.

1977-01-01

The problem of detecting a weak spectrum line, deuterium alpha, very near in wavelength to H/sub alpha/, assumed to be thousands of times as strong, is discussed from the point of view of optimizing the signal-to-noise ratio. A spectrometer consisting of three pressure scanning Fabry-Perot etalons with low reflectivity coatings was found to be the best instrument for this experiment. While no feature attributable to deuterium was detected in the planetary nebula IC 418, an upper limit relative to hydrogen of 3.4 x 10 -5 was established at the 95% confidence level. The significance of this result is discussed in light of the role played by deuterium in cosmology. The intensity ratio of the Pickering beta line (n'' = 6, n' = 4 transition of ionized helium) relative to H/sub alpha/ was measured to be 6.5 x 10 -5 . Observations of the nebular continuum made at five wavelengths in the red and near infrared are summarized and compared with predicted intensities

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

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

International Nuclear Information System (INIS)

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

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.

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

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

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

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.)

Summary of the 4th workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

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

Summary of the 4th workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

CFD for Nuclear Reactor Safety Applications (CFD4NRS-4) - Workshop Proceedings

International Nuclear Information System (INIS)

2014-01-01

Following the CFD4NRS workshops held in Garching, Germany (Sept. 2006), Grenoble, France (Sep. 2008) and Washington D.C., USA (Sept. 2010), this Workshop is intended to extend the forum created for numerical analysts and experimentalists to exchange information in the application of CFD and CMFD to NRS issues and in guiding nuclear reactor design thinking. The workshop includes single-phase and multi-phase CFD applications, and offers the opportunity to present new experimental data for CFD validation. More emphasis has been given to the experiments, especially on two-phase flow, for advanced CMFD modelling for which sophisticated measurement techniques are required. Understanding of the physics has been depen before starting numerical analysis. Single-phase and multi-phase CFD simulations with a focus on validation were performed in areas such as: single-phase heat transfer, boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These relate to NRS-relevant issues, such as pressurised thermal shock, critical heat flux, pool heat exchangers, boron dilution, hydrogen distribution in containments, thermal striping, etc. The use of systematic error quantification and the application of BPGs were strongly encouraged. Experiments providing data suitable for CFD or CMFD validation were also presented. These included local measurements using multi-sensor probes, laser-based techniques (LDV, PIV or LIF), hot-film/wire anemometry, imaging, or other advanced measuring techniques. There were over 150 registered participants at the CFD4NRS-4 workshop. The programme consisted of 48 technical papers. Of these, 44 were presented orally and 4 as posters. An additional 8 posters related to the OECD/NEA-KAERI sponsored CFD benchmark exercise on turbulent mixing in a rod bundle with spacers (MATiS-H) were presented and a special session was allocated for 6 video presentations. In addition, five keynote lectures were given by distinguished experts. The

9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

Science.gov (United States)

2010-01-01

... infectious anemia reactors and approval of laboratories, diagnostic facilities, and research facilities. 75.4... IN HORSES, ASSES, PONIES, MULES, AND ZEBRAS Equine Infectious Anemia (swamp Fever) § 75.4 Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic facilities, and...

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.

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

Present status of fusion reactor materials, 4

International Nuclear Information System (INIS)

Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

1982-01-01

Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

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.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

Science.gov (United States)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2013-01-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)

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

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well

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.

Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH4

International Nuclear Information System (INIS)

Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil; Chu, Cheun-Ho

2015-01-01

Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH 4 . Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH 4 solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH 4 during storage of NaBH 4 solution. Therefore stability of NaBH 4 and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH 4 stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH 4 stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH 4 stability and aluminun corrosion. NaBH 4 hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

Removal of FePO4 and Fe3(PO4)2 crystals on the surface of passive fillers in Fe0/GAC reactor using the acclimated bacteria

International Nuclear Information System (INIS)

Lai, Bo; Zhou, Yuexi; Yang, Ping; Wang, Juling; Yang, Jinghui; Li, Huiqiang

2012-01-01

Highlights: ► Fe 3 (PO 4 ) 2 and FePO 4 crystals would weaken treatment efficiency of Fe 0 /GAC reactor. ► Fe 3 (PO 4 ) 2 and FePO 4 crystals could be removed by the acclimated bacteria. ► FeS and sulfur in the passive film would be removed by the sulfur-oxidizing bacteria. ► Develop a cost-effective bio-regeneration technology for the passive fillers. - Abstract: As past studies presented, there is obvious defect that the fillers in the Fe 0 /GAC reactor begin to be passive after about 60 d continuous running, although the complicated, toxic and refractory ABS resin wastewater can be pretreated efficiently by the Fe 0 /GAC reactor. During the process, the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film are formed by the reaction between PO 4 3− and Fe 2+ /Fe 3+ . Meanwhile, they obstruct the formation of macroscopic galvanic cells between Fe 0 and GAC, which will lower the wastewater treatment efficiency of Fe 0 /GAC reactor. In this study, in order to remove the Fe 3 (PO 4 ) 2 and FePO 4 crystals on the surface of the passive fillers, the bacteria were acclimated in the passive Fe 0 /GAC reactor. According to the results, it can be concluded that the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film could be decomposed or removed by the joint action between the typical propionic acid type fermentation bacteria and sulfate reducing bacteria (SRB), whereas the PO 4 3− ions from the decomposition of the Fe 3 (PO 4 ) 2 and FePO 4 crystals were released into aqueous solution which would be discharged from the passive Fe 0 /GAC reactor. Furthermore, the remained FeS and sulfur (S) in the passive film also can be decomposed or removed easily by the oxidation of the sulfur-oxidizing bacteria. This study provides some theoretical references for the further study of a cost-effective bio-regeneration technology to solve the passive problems of the fillers in the zero-valent iron (ZVI) or Fe 0 /GAC reactor.

Flux distribution by neutrons semi-conductors detectors during the startup of the EL4 reactor

International Nuclear Information System (INIS)

Fuster, S.; Tarabella, A.

1967-01-01

The Cea developed neutron semi-conductors detectors which allows a quasi-instantaneous monitoring of neutrons flux distribution, when placed in a reactor during the tests. These detectors have been experimented in the EL4 reactor. The experiment and the results are presented and compared with reference mappings. (A.L.B.)

The different generation of nuclear reactors from Generation-1 to Generation-4

International Nuclear Information System (INIS)

Cognet, G.

2010-01-01

In this work author deals with the history of the development of nuclear reactors from Generation-1 to Generation-4. The fuel cycle and radioactive waste management as well as major accidents are presented, too.

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

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.

Apparent and standard molar volumes and heat capacities of aqueous Ni(ClO4)2 from 25 to 85oC

International Nuclear Information System (INIS)

Pan, P.; Campbell, A.B.

1997-01-01

Apparent molar heat capacities and volumes of aqueous Ni(ClO 4 ) 2 were measured from 25 to 85 o C over a concentration range of 0.02 to 0.8 mol-kg -1 using a Picker flow microcalorimeter and a Picker vibrating-tube densimeter. An extended Debye-Huckel equation was fitted to the experimental data to obtain expressions for the apparent molar properties as functions of ionic strength for Ni(ClO 4 ) 2 (aq). The standard-state partial molar properties for Ni(ClO 4 ) 2 (aq) in the temperature range 25 to 85 o C were obtained and can be expressed by empirical equations. The standard partial molar heat capacities and volumes for Ni 2+ (aq) from 25 to 86 o C were obtained by using the additivity rule and data for ClO - 4 (aq) in the literature. These values were extrapolated to 300 o C by employing the Helgeson-Kirkham-Flower (HKF) equations, amended to include a standard-state correction term. (author)

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

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)

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

OpenAIRE

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

2016-01-01

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

A remotely controlled CCTV system for nuclear reactor retube operations

International Nuclear Information System (INIS)

Stovman, J.A.

1984-01-01

This paper describes the CCTV Vault Observation Subsystem (VOS) under development for Ontario Hydro for the Pickering 'A' Nuclear Power Plant Large Scale Retubing program. This subsystem will be used by a supervisor and several operators to observe fuel channel replacement operations following plant shutdown and removal of the fuel bundles. VOS basically comprises 23 monochrome television camera driven circuits, a matrix switcher, 15 monitors, 9 tape recorders and 4 microphone driven sound circuits. Remote control of the camera's zoom lenses and mounts is via a digitally multiplexed control system. Design considerations include viewing requirements, reliability, radiation, redundance, and economic factors

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)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

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.

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.

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)

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

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)

Safety of research reactors. Topical issues paper no. 4

International Nuclear Information System (INIS)

Alcala-Ruiz, F.; Ferraz-Bastos, J.L.; Kim, S.C.; Voth, M.; Boeck, H.; Dimeglio, F.; Litai, D.

2001-01-01

Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety review of the research reactor facility and to verify compliance with the IAEA Safety Standards. The methods used during an INSARR mission have been collected and analysed. Some of the important issues identified are the following: general ageing of the facility; uncertain status of many research reactors (in extended shutdown); indefinite deferral of return to operation or decommissioning; inadequate regulatory supervision; insufficient systematic (periodic) reassessment of safety; lack of quality assurance (QA) programmes; lack of an international safety convention or arrangement; lack of financial support for safety measures (e.g. safety reassessment, safety upgrading, decommissioning) and utilization; lack of clear utilization programmes; inadequate emergency preparedness; inadequate safety documentation (e.g. safety analysis report, operating rules and procedures, emergency plan); inadequate funding of shutdown reactors; weak safety culture; loss of expertise and corporate memory; loss of information concerning radioactive materials contained in retired experimental devices stored in the facility indefinitely; obsolescence of equipment and lack of spare parts; inadequate training and qualifications of regulators and operators; safety implications of new fuel types. These issues have been addressed by the IAEA Secretariat and the chairman of the International Nuclear Safety Advisory Group (INSAG). INSAG has identified three major safety issues that are: the increasing age of research reactors, the number of research reactors that are not operating anymore but have not been decommissioned, and the number of research reactors in countries that do not have appropriate regulatory authorities. This issue paper discusses the concerns generated by an analysis of the results of INSARR missions and those expressed by INSAG. The

Study of cold neutron sources: Implementation and validation of a complete computation scheme for research reactor using Monte Carlo codes TRIPOLI-4.4 and McStas

International Nuclear Information System (INIS)

Campioni, Guillaume; Mounier, Claude

2006-01-01

The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

International Nuclear Information System (INIS)

2007-03-01

In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

International Nuclear Information System (INIS)

2009-01-01

In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-02-15

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH{sub 4}

Energy Technology Data Exchange (ETDEWEB)

Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil [Sunchon National University, Suncheon (Korea, Republic of); Chu, Cheun-Ho [ETIS Co, Gimpo (Korea, Republic of)

2015-12-15

Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH{sub 4}. Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH{sub 4} solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH{sub 4} during storage of NaBH{sub 4} solution. Therefore stability of NaBH{sub 4} and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH{sub 4} stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH{sub 4} stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH{sub 4} stability and aluminun corrosion. NaBH{sub 4} hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

Interactions of RuO4(g) with different surfaces in nuclear reactor containments

International Nuclear Information System (INIS)

Holm, J.; Glaenneskog, H.; Ekberg, C.

2008-07-01

During a severe nuclear reactor accident with air ingress, ruthenium in the form of RuO4 can be released from the nuclear fuel. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. This work has investigated the distribution of RuO4 between an aqueous and gaseous phase in the temperature interval of 20-50 deg. C by on-line measurements with an experimental set-up made of glass. The experiments showed that RuO4 is almost immediately distributed in the aqueous phase after its introduction in the set-up in the entire temperature interval. However, the deposition of ruthenium on the glass surfaces in the system was significant. The speciation of the ruthenium on the glass surfaces was studied by SEM-EDX and ESCA and was determined to be the expected RuO2. Experiments of interactions between gaseous ruthenium tetroxide and the metals aluminium, copper and zinc have been investigated. The metals were treated by RuO4 (g) at room temperature and analyzed with ESCA, SEM and XRD. The analyses show that the black ruthenium deposits on the metal surfaces were RuO2, i.e. the RuO4 (g) has been transformed on the metal surfaces to RuO2(s). The analyses showed also that there was a significant deposition of ruthenium tetroxide especially on the copper and zinc samples. Aluminium has a lower ability to deposit gaseous ruthenium tetroxide than the other metals. The conclusion that can be made from the results is that surfaces in nuclear reactor containments will likely reduce the source term in the case of a severe accident in a nuclear power plant. (au)

Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

2010-01-01

The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

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)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

2007-01-01

CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

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.)

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

International Nuclear Information System (INIS)

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

The EL-4 reactor. Changing of a pressure tube on a test loop

International Nuclear Information System (INIS)

Foulquier, H.; Clara, P.

1964-01-01

Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

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%

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

International Nuclear Information System (INIS)

2012-03-01

In the fiscal year 2007, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) and the JRR-4 was operated for 92 days. JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 262 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analyses, prompt gamma-ray analyses, and others submitted by the users in JAEA and other Organizations. (author)

Canadian supercritical water reactor modeling using G4STORK

International Nuclear Information System (INIS)

Ford, W.; Buijs, A.

2015-01-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Canadian supercritical water reactor modeling using G4STORK

Energy Technology Data Exchange (ETDEWEB)

Ford, W.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

2015-07-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2012-01-01

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

2012-01-15

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Operating reactors licensing actions summary. Volume 4, No. 9

International Nuclear Information System (INIS)

1984-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Vol. 4, No. 2

International Nuclear Information System (INIS)

1984-04-01

This 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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1985-07-01

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

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.

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

The Chernobyl-4 Reactor and the possible causes of the accident

International Nuclear Information System (INIS)

Motte, F.

1986-01-01

A description and information about the Chernobyl nuclear reactor is given. Some comparison elements between the RBMK reactor type and GCR, CANDU, SGHWR and Hanford N reactor types are presented. A scenario of the possible causes of the accident is discussed. (A.F.)

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

Present status of decommissioning in the Musashi Reactor Facility (4)

International Nuclear Information System (INIS)

Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

2012-01-01

The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

Calculations of fuel burn up and radionuclide inventories in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor as a function of the reactor operating time for 10, 20, and 30 k W operating power levels. The uranium burn up rate and burn up percentage, the amounts of the plutonium isotopes, the concentrations and radioactivities of the fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well. The CITATION code is used to calculate the changes in the effective multiplication factor of the reactor.(author)

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)

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

A simulation of a pebble bed reactor core by the MCNP-4C computer code

Directory of Open Access Journals (Sweden)

Bakhshayesh Moshkbar Khalil

2009-01-01

Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

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

Nuclear reactor technology progress report, vol. 4

International Nuclear Information System (INIS)

1981-01-01

The works of the Engineering Section, Fast Experimental Reactor Division, are roughly classified into the technologies concerning the reactor core, abnormality monitoring, the plant, purity control and operation planning. In this paper, the activities of the Engineering Section, the operational results of Joyo and the foreign informations on FBRs in this quarter are reported. The second regular inspection carried out successively from the previous quarter was completed, and the fourth cycle operation of Joyo at 75 MW was started. The measurement of CP around the primary system pipings and equipments, the preliminary test of a core flow meter for Monju, and the various characteristic tests were carried out during this period. 2 N reports, 1 SA report and 63 memos were drawn up in this quarter. The test plan to be carried out during the period of the fourth to sixth cycle operations in this last year using the MK-1 core was formed and decided. Various meetings within and outside the division are reported. The data obtained in the operational characteristic test and special test are shown. As the results concerning the reactor technologies, the development of dosimetry techniques, the measurement and analysis of the core characteristics, the measurement of the temperature and flow velocity of coolant at the fuel assembly exit, the system pressure loss in the primary cooling system and others are reported. (Kako, I.)

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

Energy Technology Data Exchange (ETDEWEB)

Guentay, S [ed.; Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs.

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

International Nuclear Information System (INIS)

Guentay, S.

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs

The accident of Chernobylsk-4 reactor and its consequences

International Nuclear Information System (INIS)

1986-01-01

This report deals with the particulars of the accident as communicated by the Soviet delegation at an IAEA meeting by the and of August 1986. It was stated that the consequences emanated from the inherent instability of the design of the reactor, the deviation from the safety rules by the operators and the lack of a sight reactor containment. (G.B.)

Magnetic susceptibility as a direct measure of oxidation state in LiFePO4 batteries and cyclic water gas shift reactors.

Science.gov (United States)

Kadyk, Thomas; Eikerling, Michael

2015-08-14

The possibility of correlating the magnetic susceptibility to the oxidation state of the porous active mass in a chemical or electrochemical reactor was analyzed. The magnetic permeability was calculated using a hierarchical model of the reactor. This model was applied to two practical examples: LiFePO4 batteries, in which the oxidation state corresponds with the state-of-charge, and cyclic water gas shift reactors, in which the oxidation state corresponds to the depletion of the catalyst. In LiFePO4 batteries phase separation of the lithiated and delithiated phases in the LiFePO4 particles in the positive electrode gives rise to a hysteresis effect, i.e. the magnetic permeability depends on the history of the electrode. During fast charge or discharge, non-uniform lithium distributionin the electrode decreases the hysteresis effect. However, the overall sensitivity of the magnetic response to the state-of-charge lies in the range of 0.03%, which makes practical measurement challenging. In cyclic water gas shift reactors, the sensitivity is 4 orders of magnitude higher and without phase separation, no hysteresis occurs. This shows that the method is suitable for such reactors, in which large changes of the magnetic permeability of the active material occurs.

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.)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

2009-06-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

2012-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2010 and March 31, 2011. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for research reactors and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

Generalities about nuclear reactors

International Nuclear Information System (INIS)

Jaouen, C.; Beroux, P.

2012-01-01

From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

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.

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

Detection of pressure tube leaks relying on moisture beetles only

International Nuclear Information System (INIS)

Kenchington, J.M.; Choi, A.; Jin, Y.

2004-01-01

A major decision was made for Pickering NGS A Annulus Gas System (ACS) that detection of a pressure tube (PT) leak should be achieved by using only moisture beetles and that dew point monitors would provide 'early warning' without status to shut down the reactor. Experience with Unit 3 has shown that dew point monitoring of pressure tube leaks was particularly subject to gas leaks and surface adsorption effects. Unit 4 was the first one to be converted during the full scale pressure tube replacement programme. Because of the fundamental change in design philosophy, moisture injection tests were carried out during commissioning to demonstrate that performance matched design. In particular it was necessary to show that leak before break (LBB) would be achieved if a leak occurred in the limiting string. Units 1 and 3 have since been converted. No decision has been taken to convert Pickering B units as gas leaks are small and no significant adsorption effects are anticipated. Hence dew point monitoring will not be impaired. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

2013-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

An initial assessment of the Chernobyl-4 reactor accident release source

International Nuclear Information System (INIS)

Macdonald, H.F.; ApSimon, H.M.; Wilson, J.J.N.

1986-07-01

The long-range atmospheric dispersion model MESOS has been used to provide a preliminary evaluation of the effects over Western Europe of radioactivity released during the accident which occurred at the Chernobyl-4 reactor in the USSR in April 1986. The results of this analysis have been compared with observations during the first week or so following the accident of airborne contamination levels at a range of locations across Europe in order to obtain an estimate of accident release source. The work presented here was performed during the 6-8 weeks following the accident and the results obtained will be subject to refinement as more detailed data become available. However, at this early stage they indicate a release source for the Chernobyl accident, expressed as a fraction of the estimated reactor core inventory, of approx. 15-20% of the iodine and caesium isotopes, approx. 1% of the ruthenium and lesser amounts of the other fission products and actinides, together with an implied major fraction of the krypton and xenon noble gases. (author)

Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

International Nuclear Information System (INIS)

1980-01-01

The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

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)

Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

International Nuclear Information System (INIS)

Khattab, K.; Omar, H.; Ghazi, N.

2009-01-01

A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation

Directory of Open Access Journals (Sweden)

Subagjo

2014-07-01

Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.

Evaluation of the OSCAR-4/MCNP calculation methodology for radioisotope production in the SAFARI-1 reactor

International Nuclear Information System (INIS)

Karriem, Z.; Zamonsky, O.M.

2014-01-01

The South African Nuclear Energy Corporation SOC Ltd (Necsa) is a state owned nuclear facility which owns and operates SAFARI-1, a 20 MW material testing reactor. SAFARI-1 is a multi-purpose reactor and is used for the production of radioisotopes through in-core sample irradiation. The Radiation and Reactor Theory (RRT) Section of Necsa supports SAFARI-1 operations with nuclear engineering analyses which include core-reload design, core-follow and radiation transport analyses. The primary computer codes that are used for the analyses are the OSCAR-4 nodal diffusion core simulator and the Monte Carlo transport code MCNP. RRT has developed a calculation methodology based on OSCAR-4 and MCNP to simulate the diverse in-core irradiation conditions in SAFARI-1, for the purpose of radioisotope production. In this paper we present the OSCAR-4/MCNP calculation methodology and the software tools that were developed for rapid and reliable construction of MCNP analysis models. The paper will present the application and accuracy of the methodology for the production of yttrium-90 ( 90 Y) and will include comparisons between calculation results and experimental measurements. The paper will also present sensitivity analyses that were performed to determine the effects of control rod bank position, representation of core depletion state and sample loading configuration, on the calculated 90 Y sample activity. (author)

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-03-01

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

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

International Nuclear Information System (INIS)

El-Messeiry, A.M.

1996-01-01

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

Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Lee, Y.K.; Hugot, F.X.

2011-01-01

The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

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

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2011-01-01

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system

Energy Technology Data Exchange (ETDEWEB)

Kuscu, Ozlem Selcuk, E-mail: oselcuk@mmf.sdu.edu.tr [Department of Environmental Engineering, Engineering and Architecture Faculty, Sueleyman Demirel University, Cuenuer Campus, 32260 Isparta (Turkey); Sponza, Delia Teresa [Dokuz Eyluel University, Engineering Faculty, Environmental Engineering Department, Buca Kaynaklar campus, Izmir (Turkey)

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system.

Science.gov (United States)

Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR. Copyright © 2011 Elsevier B.V. All rights reserved.

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%.

Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

International Nuclear Information System (INIS)

Van Heek, A.I.

1996-06-01

Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

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.)

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-04-01

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

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

Problems of creating fuel elements for fast gas-cooled reactors working on N2O4-dissociating coolant

International Nuclear Information System (INIS)

Nesterenko, V.B.; Zelensky, V.F.; Kolykhan, L.I.; Karpenko, G.V.; Krasnorutsky, V.S.; Isakov, V.P.; Ashikhmin, V.P.; Permyakov, L.N.

1985-01-01

A variant of fast gas-cooled reactors is one using dissociating N 2 O 4 nitrogen tetroxide as a coolant. This type of reactors is promising because of great thermal effects of dissociation reactions while heating and recombination while cooling; small latent heat of evaporation; high heat transfer coefficient owing to additional heat transfer in a chemical reaction; high N 2 O 4 density in a gas state at operation parameters. The mentioned advantages give possibility to create a small turbine, heat exchange apparatus and to get high heat production in the active zone. All this opens new ways to increase power plants effectiveness

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.

Medical aspects of the nuclear accident in the Chernobylsk-4 reactor

International Nuclear Information System (INIS)

Arndt, D.; Schmidt, W.

1989-01-01

The Kiev conference on the Chernobylsk reactor accident was concerned with the following items: (1) Medical consequences and organization of medical assistance as well as aftercare of radiation-exposed persons. (2) Analysis of the postirradiation situation and judgement of the consequences of the accident as to the USSR population. (3) Peculiarities of external and internal radiation exposure of the population in the area controlled. (4) Organization and efficiency of the epidemiological register of the USSR. (5) Organization and judgement of educational work and public relations concerning the sanitary conditions in populations exposed to an increased contamination

Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

Elements on reactor control

International Nuclear Information System (INIS)

Bruna, G.B.

1998-01-01

In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

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

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

International Nuclear Information System (INIS)

2012-12-01

The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

The development of octagon Zr-4 alloy tube for heating reactors

International Nuclear Information System (INIS)

Yang Fanglin; Yang Yingli; Wang Guangshen

1989-10-01

The asymmetrical octagon Zr-4 alloy tubes which are used for fuel assembly in the heating reactor have been developed. The thickness of tube wall is 1.5 mm and the length is 1725 mm. The long side of the octagon is 138.7 0.3 +0.2 mm, the short side is 93.1 ± 0.1 mm. To manufacture these tubes a stretch draw forming processing method is adopted. The process is divided into two phases. In the first phase, a short draw mould is used to stretch the Zr-4 alloy tube. In the second phase, a long draw mould, its length is equal to the end-produt length, is used to complete the final processing. The size accuracy and repeatability of this method are excellent and can fully meet the design requirements

Annex VII - Diagrams: 1. Reactor operation (1960-1977); 2. Mean daily reactor power density in 1977; 3. Monthly reactor power for 1977; 4. percent of utilization of experimental space in 1977; Prilog VII - Dijagrami: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-12-15

This Annex includes the following diagrams: 1. Annual Reactor RA power production (MWh) for the period from 1960-1977; 2. Mean daily reactor power density MW in 1977; 3. Monthly reactor power production (MWh) for 1977; 4. percent of utilization of experimental space in 1977. [Serbo-Croat] Ovaj prilog sadrzi dijagrame: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

2014-03-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

The molten salt reactor adventure

International Nuclear Information System (INIS)

MacPherson, H.G.

1985-01-01

A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF 4 -ThF 4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-08-01

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 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang 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 Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

Management of spent fuel from research reactors - Brazilian progress report (within the framework of Regional Project IAEA-RLA-4/018)

International Nuclear Information System (INIS)

Soares, A.J.; Silva, J.E.R.

2005-01-01

There are four research reactors in Brazil. For three of them, because of the low reactor power and low burn-up of the fuel, except for the concern about ageing, spent fuel storage is not a problem. However for one of the reactors, more specifically IEA-R1 research reactor, the storage of spent fuel is a major concern, because, according to the proposed operation schedule for the reactor, unless an action is taken, by the year 2009 there will be no more racks available to store its spent fuel. This paper gives a brief description of the type and amount of fuel elements utilized in each one of the Brazilian research reactors, with a short discussion about the storage capacity at each installation. It also gives a description of the activities developed by Brazilian engineers and researchers during the period between 2001 and 2004, within the framework of regional project 'RLA-4/018-Management of Spent Fuel from Research Reactors'. As a conclusion, we can say that the advances of the project, and the integration promoted among the engineers and researchers of the participant countries were of fundamental importance for Brazilian researchers and engineers to understand the problems related to the storage of spent fuel, and to make a clear definition about the most suitable alternatives for interim storage of the spent fuel from IEAR1 research reactor. (author)

What occurred in the reactors

International Nuclear Information System (INIS)

Kudo, Kazuhiko

2013-01-01

Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

International Nuclear Information System (INIS)

Bowman, S.M.; Suto, T.

1996-10-01

ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

Fast reactors

International Nuclear Information System (INIS)

Vasile, A.

2001-01-01

Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

International Nuclear Information System (INIS)

Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

2017-01-01

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

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

RA-0 reactor. New neutronic calculations; Reactor RA-0. Nuevos calculos neutronicos

Energy Technology Data Exchange (ETDEWEB)

Rumis, D; Leszczynski, F

1991-12-31

An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core`s interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author). [Espanol] En este trabajo se actualizan los calculos neutronicos realizados para el reactor RA-0, instalado en la Facultad de Ciencias Exactas, Fisicas y Naturales de la Universidad Nacional de Cordoba. Se describen los calculos realizados hasta la fecha y los resultados obtenidos. Las tecnicas incorporadas al calculo de un reactor como el RA-0 permiten predecir en detalle el comportamiento del flujo en el interior del nucleo y en el reflector, lo que sera una importante ayuda en el diseno de experimentos. En particular, el empleo del codigo WIMSD4 para calculos del reactor completo constituye una novedad en las posibles aplicaciones de ese codigo para resolver problemas que se presentan en la practica. (Autor).

Delayed hydrogen cracking of zirconium alloy pressure tubes

International Nuclear Information System (INIS)

Jackman, A.H.; Dunn, J.T.

1976-10-01

After several years of almost continuous service, Pickering Units 3 and 4 have both experienced long outages to replace cracked pressure tubes. This report summarizes the status of the investigation into the cause of the cracks as of May 1976. The basic cause of the cracking was the presence of very high residual tensile stresses in the pressure tubes due to improper rolling procedures. These residual stresses are being reduced to acceptable levels by local stress relieving techniques at Bruce G.S. and in future reactors improvements in rolling procedures and changes in pressure tube specifications will prevent a recurrence of this problem. (author)

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)

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

The IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2001-01-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2001-04-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1976-09-01

Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

Studies of conceptual spheromak fusion reactors

International Nuclear Information System (INIS)

Katsurai, M.; Yamada, M.

1982-01-01

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

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...

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

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)

2001-02-01

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 4th 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 Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

Hythane (H2 and CH4) production from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals via up-flow anaerobic self-separation gases reactor

International Nuclear Information System (INIS)

Mahmoud, Mohamed; Elreedy, Ahmed; Pascal, Peu; Sophie, Le Roux; Tawfik, Ahmed

2017-01-01

Highlights: • Bio-hythane production from polyester wastewater via UASG reactor was assessed. • Impacts of influent contamination by 1,4-dioxane and heavy metals were discussed. • Maximum volumetric H 2 and CH 4 productions of 0.12 and 1.06 L/L/d were achieved. • Significant drop in CH 4 production was resulted at OLR up to 1.07 ± 0.06 gCOD/L/d. • Bioenergy recovery through UASG economically achieved a net profit of 10,231 $/y. - Abstract: A long-term evaluation of hythane generation from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals was investigated in a continuous up-flow anaerobic self- separation gases (UASG) reactor inoculated with mixed culture. The reactor was operated at constant hydraulic retention time (HRT) of 96 h and different organic loading rates (OLRs) of 0.31 ± 0.04, 0.71 ± 0.08 and 1.07 ± 0.06 gCOD/L/d. Available data showed that volumetric hythane production rate was substantially increased from 0.093 ± 0.021 to 0.245 ± 0.016 L/L/d at increasing OLR from 0.31 ± 0.04 to 0.71 ± 0.08 gCOD/L/d. However, at OLR exceeding 1.07 ± 0.06 gCOD/L/d, it was dropped to 0.114 ± 0.016 L/L/d. The reactor achieved 1,4-dioxane removal efficiencies of 51.8 ± 2.8, 35.9 ± 1.6 and 26.3 ± 1.6% at initial 1,4-dioxane concentrations of 1.14 ± 0.28, 1.97 ± 0.41 and 4.21 ± 0.30 mg/L, respectively. Moreover, the effect and potential removal of the contaminated by heavy metals (i.e., Cu 2+ , Mn 2+ , Cr 3+ , Fe 3+ and Ni 2+ ) were highlighted. Kinetic modelling and microbial community dynamics were studied, according to each OLR, to carefully describe the UASG performance. The economic analysis showed a stable operation for the anaerobic digestion of unsaturated polyester resin wastewater using UASG, and the maximum net profit was achieved at OLR of 0.71 ± 0.08 gCOD/L/d.

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-12-15

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10{sup -3}) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

International Nuclear Information System (INIS)

2013-01-01

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10 -3 ) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the Proceedings of

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

The Integral Fast Reactor

International Nuclear Information System (INIS)

Chang, Y.I.

1988-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. 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. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Factors affecting biological reduction of CO{sub 2} into CH{sub 4} using a hydrogenotrophic methanogen in a fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Hyung; Pak, Daewon [Seoul National University of Science and Technology, Seoul (Korea, Republic of); Chang, Won Seok [Korea District Heating Corp, Seongnam (Korea, Republic of)

2015-10-15

Biological conversion of CO{sub 2} was examined in a fixed bed reactor inoculated with anaerobic mixed culture to investigate influencing factors, the type of packing material and the composition of the feeding gas mixture. During the operation of the fixed bed reactor by feeding the gas mixture (80% H{sub 2} and 20% CO{sub 2} based on volume basis), the volumetric CO{sub 2} conversion rate was higher in the fixed bed reactor packed with sponge due to its large surface area and high mass transfer from gas to liquid phase compared with PS ball. Carbon dioxide loaded into the fixed bed reactor was not completely converted because some of H{sub 2} was used for biomass growth. When a mole ratio of H{sub 2} to CO{sub 2} in the feeding gas mixture increased from 4 to 5, CO{sub 2} was completely converted into CH{sub 4}. The packing material with large surface area is effective in treating gaseous substrate such as CO{sub 2} and H{sub 2}. H{sub 2}, electron donor, should be providing more than required according to stoichiometry because some of it is used for biomass growth.

Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

Energy Technology Data Exchange (ETDEWEB)

Johnson, A.B. Jr.

1976-09-01

Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

International Nuclear Information System (INIS)

Dupleac, D.; Perez, M.; Reventos, F.; Allison, C.

2011-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

Energy Technology Data Exchange (ETDEWEB)

Dupleac, D., E-mail: danieldu@cne.pub.ro [Politehnica Univ. of Bucharest (Romania); Perez, M.; Reventos, F., E-mail: marina.perez@upc.edu, E-mail: francesc.reventos@upc.edu [Technical Univ. of Catalonia (Spain); Allison, C., E-mail: iss@cableone.net [Innovative Systems Software (United States)

2011-07-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

PERFORMA NEUTRONIK BAHAN BAKAR LiF-BeF2-ThF4-UF4 PADA SMALL MOBILE-MOLTEN SALT REACTOR

Directory of Open Access Journals (Sweden)

S. N. Rokhman

2015-04-01

Full Text Available Telah dilakukan analisis terhadap performa neutronik bahan bakar garam lebur LiF-BeF2-ThF4-UF4 pada Small Mobile-Molten Salt Reactor (SM-MSR. Penyesuaian konfigurasi teras dan temperatur operasi harus dilakukan untuk penggunaan bahan bakar baru tersebut agar mencapai keff > 1 dan CR (conversion ratio > 1 pada fraksi 0,5% 233U, 20% 232Th, 28% Li, 51,5% Be. Setelah didapat nilai keff ≈ 1 dan CR ≈ 1, dilakukan analisis pengaruh perubahan Th terhadap Be dan Be terhadap Li yang terlihat dalam perubahan parameter keff dan CR. Setelah itu fraksi 233U divariasi antara 0,5–0,46% untuk memperoleh keff > 1 dan CR > 1. Dalam perhitungan koefisien reaktifitas temperatur (αT, temperatur teras dinaikkan sebesar +25K dan +50K., dan untuk koefisien reaktifitas void (αV, densitas bahan bakar dikurangi hingga 90%. Hasil perhitungan menunjukkan bahwa pengurangan Th terhadap Be menyebabkan penurunan nilai CR dan naiknya keff akibat berkurangnya material fertil. Sebaliknya penambahan Be terhadap Li mengakibatkan terjadi kenaikan nilai keff dan menurunkan CR, akibat laju serapan Li lebih besar dari Be. Pada 5 (lima fraksi 233U dalam rentang 0,5–0,49%, hasil perhitungan keff dan CR masing-masing bervariasi dalam rentang 1,00001 - 1,00327 dan 1,00016 - 1,00731. Untuk faktor puncak daya (PPF, hasil perhitungan memberikan nilai dalam rentang 2,4311 -2,4714. Sedangkan untuk parameter keselamatan, koefisien reaktivitas temperatur (αT dan reaktivitas void (αV masingmasing bernilai negatif dalam rentang 4,972×10-5 - 5,909×10-5 dan 2,596×10-2- 2,8287×10-2 ∆k/k/K. Dapat disimpulkan bahwa teras SM-MSR memberikan nilai negatif di kedua koefisien reaktivitas tersebut untuk setiap fraksi,, sehingga memenuhi kriteria keselamatan dan keselamatan melekat. Kata kunci: SM-MSR (small mobile-molten salt reactor, bahan bakar LiF-BeF2-ThF4-UF4, keselamatan melekat, koefisien reaktivitas temperatur, koefisien reaktivitas void   The analysis of neutronic performance has

Spatial kinetics in nuclear reactor systems. Chapter 4

International Nuclear Information System (INIS)

Owens, D.H.

1980-01-01

The problem of constructing a low-order linear lumped-parameter model of xenon-induced spatial power oscillations in a large, cylindrical nuclear power reactor to replace an (assumed known) nonlinear distributed parameter model is examined. Model expansion and finite difference methods are used together to provide a successful solution to the problem. (U.K.)

Synthesis of superior fast charging-discharging nano-LiFePO4/C from nano-FePO4 generated using a confined area impinging jet reactor approach.

Science.gov (United States)

Liu, Xiao-min; Yan, Pen; Xie, Yin-Yin; Yang, Hui; Shen, Xiao-dong; Ma, Zi-Feng

2013-06-14

LiFePO4/C nanocomposites with excellent electrochemical performance is synthesized from nano-FePO4, generated by a novel method using a confined area impinging jet reactor (CIJR). When discharged at 80 C (13.6 Ag(-1)), the LiFePO4/C delivers a discharge capacity of 95 mA h g(-1), an energy density of 227 W h kg(-1) and a power density of 34 kW kg(-1).

Nuclear data usage for research reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

1996-01-01

In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Proceedings of the 4th international symposium on material testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Suzuki, Masahide

2012-03-01

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Proceedings of the 4th international symposium on material testing reactors

Energy Technology Data Exchange (ETDEWEB)

Ishihara, Masahiro; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Analysis of gamma dose for 4,8 gU/cm3 density silicide core at the RSG-GAS reactor using MCNP code

International Nuclear Information System (INIS)

Ardani

2011-01-01

Radiation safety analysis should be done following of substitution of fuel density of 2.96 gU/cc to density of 4,8 gU/cc silicide fuels for the RSG-GAS reactor. MCNP-5 code has been used to perform gamma dose calculation of the RSG-GAS reactor. Gamma radiation source at reactor consists of capture gamma rays, prompt fission gamma rays, and gamma rays of decay of fission and activation products. The strength of the prompt fission gamma rays is obtained by gamma releases of fission process of U-235 and reactor power of 30 MWt., during 46,6 days operation. Radiation dose is calculated at the experimental hall by detection point at the surface of outer of biological shielding and the operation hall by detection point at the top of the pool. The calculation is conducted at reactor on the normal operation and on the worst postulated accident causing the water level at the pool decreases. Calculation result shows that the biggest source strength of gamma rays come from the decay process. The highest calculated dose at the experiment hall is 4,07x10 -3 μSv/h, far from the maximum external dose permitted 25 μSv/h. The highest calculated dose at the operation hall is 19.98 μSv/h. Even though the calculated dose is still acceptable but this is close to the maximum permitted dose for worker. It concluded that loading of 4,8 gU/cc silicide fuel for the RSG-GAS still safe. (author)

Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

Directory of Open Access Journals (Sweden)

SA Hosseini

2013-09-01

Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

Energy Technology Data Exchange (ETDEWEB)

Gomez-De Jesus, A.; Romano-Baez, F.J.; Leyva-Amezcua, L.; Juarez-Ramirez, C.; Ruiz-Ordaz, N. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico); Galindez-Mayer, J. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico)], E-mail: cmayer@encb.ipn.mx

2009-01-30

For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A{sub D}/A{sub R} ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values {epsilon}{sub G}, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ({approx}95%), and by the stoichiometric release of chloride ions from the halogenated compound ({approx}80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

Radiation protection at the RA Reactor in 1993, RA research reactor, Part

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

1993-01-01

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 and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

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

Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

2007-04-15

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 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin 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 Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Cermet fuel reactors

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

International Nuclear Information System (INIS)

Jammes, Ch.

2010-07-01

After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

The integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Yoon I.; Marchaterre, J.F.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. 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. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

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

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Mirror hybrid reactor optimization studies

International Nuclear Information System (INIS)

Bender, D.J.

1976-01-01

A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

Energy Technology Data Exchange (ETDEWEB)

Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M. [National Research Centre “Kurchatov Institute,”, Konstantinov Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I. [National Research Centre “Kurchatov Institute,” (Russian Federation); Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K. [State Scientific Centre Research Institute of Atomic Reactors (Russian Federation); and others

2015-10-15

In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Cupac, S.; Stanic, A.

1990-01-01

RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

Molten salt reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

2014-01-01

Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

Recent fuel handling experience in Canada

International Nuclear Information System (INIS)

Welch, A.C.

1991-01-01

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

Possibilities of TWR and long life reactor

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Shimazu, Yoichiro; Handa, Norihiko

2010-01-01

Bill Gates identified the need to switch to zero-emission energy and clarified investing in Terra Power developing the TWR (Traveling Wave Reactor) in February 2010. He also visited Toshiba developing small reactor 4S (Super Safe Small and Simple). In Japan design studies of the TWR have been conducted on the CANDLE reactor without refueling and the 4S long life reactor with maintenance free. In this feature article, the state of R and D on the TWR in Japan and IAEA's activities on small reactors without online refueling were reviewed in addition to articles on impacts of Bill Gates' investment in the TWR and state of the TWR development from an interview with John Gilleland of Terra Power. (T. Tanaka)

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.

Public information circular for shipments of irradiated reactor fuel. Revision 4

International Nuclear Information System (INIS)

1984-06-01

This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

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

International Nuclear Information System (INIS)

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

1982-12-01

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

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2014. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Osa, Akihiko; Imahashi, Masaki; Hirane, Nobuhiko; Motome, Yuiko; Tayama, Hidekazu; Tamura, Itaru; Harada, Yuko; Sakata, Mami; Kadokura, Masakazu; Takita, Chiharu

2017-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3 (Japan Research Reactor No.3), JRR-4 (Japan Research Reactor No.4), NSRR (Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2014. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration, and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Kashima, Yoichi; Murayama, Yoji; Nakamura, Kiyoshi; Uno, Yuki; Hirane, Nobuhiko; Ohuchi, Hitoshi; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Harada, Yuko; Kadokura, Masakazu; Machi, Sumire; Takita, Chiharu

2015-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2013. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

CO2 photoreduction using NiO/InTaO4 in optical-fiber reactor for renewable energy

NARCIS (Netherlands)

Wang, Zhen-Yi; Chou, Hung-Chi; Wu, Jeffrey C.S.; Tsai, Din Ping; Mul, Guido

2010-01-01

The photocatalytic reduction of CO2 into fuels provides a direct route to produce renewable energy from sunlight. NiO loaded InTaO4 photocatalyst was prepared by a sol–gel method. Aqueous-phase CO2 photoreduction was performed in a quartz reactor to search for the highest photoactivity in a series

Kinetic analysis of sub-prompt-critical reactor assemblies

International Nuclear Information System (INIS)

Das, S.

1992-01-01

Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

International Nuclear Information System (INIS)

Pierre, J.R.M.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

Turning points in reactor design

International Nuclear Information System (INIS)

Beckjord, E.S.

1995-01-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

Turning points in reactor design

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-09-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

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

Energy Technology Data Exchange (ETDEWEB)

1975-07-01

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

Introduction of advanced pressurized water reactors in France

International Nuclear Information System (INIS)

Millot, J.P.; Nigon, M.; Vitton, M.

1988-01-01

Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

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

Fast reactor irradiation effects on fracture toughness of Si_3N_4 in comparison with MgAl_2O_4 and yttria stabilized ZrO_2

International Nuclear Information System (INIS)

Tada, K.; Watanabe, M.; Tachi, Y.; Kurishita, H.; Nagata, S.; Shikama, T.

2016-01-01

Fracture toughness of silicon nitride (Si_3N_4), magnesia-alumina spinel (MgAl_2O_4) and yttria stabilized zirconia (8 mol%Y_2O_3–ZrO_2) was evaluated by the Vickers-indentation technique after the fast reactor irradiation up to 55 dpa (displacement per atom) at about 700 °C in the Joyo. The change of the fracture toughness by the irradiation was correlated with nanostructural evolution by the irradiation, which was examined by transmission electron microscopy. The observed degradation of fracture toughness in Si_3N_4 is thought to be due to the relatively high density of small-sized of the irradiation induced defects, which should be resulted from a large amount of transmutation gases of hydrogen and helium. Observed improvement of fracture toughness in MgAl_2O_4 was due to the blocking of crack propagation by the antiphase boundaries. The radiation effects affected the fracture toughness of yttria stabilized zirconia at 55 dpa, suggesting that the generated high density voids would affect the propagation of cracks. - Highlights: • Si_3N_4, MgAl_2O_4 and YSZ were neutron irradiated up to 55dpa around 700 °C in the Joyo. • They are candidate ceramics for the inert matrices of nuclear fuels in the fast reactors. • The irradiation enhanced the fracture toughness of MgAl_2O_4 and YSZ, while degraded that of Si_3N_4. • The toughness changes were correlated with radiation induced defects and transmutation gases.

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

Reactor Physics Behind the Chernobyl Accident

International Nuclear Information System (INIS)

Reisch, F.

1999-01-01

There are some fourteen Chernobyl type of power reactors (1000 MWe) in operation at five different sites in Eastern Europe. In Russia; in St. Petersburg (4). in Smolensk (3). and in Kursk (4) in the Ukraine in Chernobyl (l) and in Lithuania in Ignalina (2). The oldest one is west of St. Petersburg and the most powerful one is in Ignalina. The reactors at St. Petersburg and in Lithuania are near to the Baltic sea. An intricate reactor construction was the most important cause of the accident. There were other reasons too: human error. politics and economics

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

Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

International Nuclear Information System (INIS)

2001-01-01

Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

African Zoology - Vol 23, No 4 (1988)

African Journals Online (AJOL)

Book Reviews · EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT · DOWNLOAD FULL TEXT DOWNLOAD FULL TEXT. Lynne Whiffler, S.G. Compton, M.R. Perrin, M.D. Picker, M.R. Perrin, P.C. Magnuson, Jay O'Keefe, A.N. Hodgson, P Hewitt, S Endrödy-Younga, L.G. Underhill, H.M. Dott, R.T.F. Bernard, C.R. Brown, 370- ...

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

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

1989-01-01

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

Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

DEFF Research Database (Denmark)

Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

2013-01-01

Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

Life extension, power upgrade, and return to service work for Pickering NGS and other PWR and CANDU plants

International Nuclear Information System (INIS)

Millman, J.; Idvorian, N.; Schneider, W.

2002-01-01

Work on life extension, power upgrade and return to service has been performed and is in progress for a number of PWR and CANDU plants. For PWR plants, power upgrade work has been done for the new replacement steam generators in several cases. This work consists of redoing the formal equipment qualification analysis and reports for the uprated operating conditions to support the application for license adjustment. Life extension assessments have been performed for several CANDU plants. These are highly detailed assessments in which the particular steam generator is reassessed part by part as to the ability of each to sustain full life operation and also extended life operation. Return to service work for Pickering NGSA specifically has included this type of assessment and also specific repair, cleaning and retrofit activities including secondary side inspection, waterlancing, divider plate repair, eddy current inspection, etc. Steam generator modifications and retrofit work have been performed in a number of cases. The paper discusses various life extension, power upgrade, equipment modification and return to service activities all of which are part of the renewed drive in the industry to realise the full potential of nuclear plants by getting more and better performance from the extended service of existing plants. (author)

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Astrid (fast breeder nuclear reactor)

International Nuclear Information System (INIS)

2014-01-01

This document presents ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a French project of sodium-cooled fast breeder reactor, fourth generation reactor which should be fuelled by uranium 238 rather than uranium 235, and should therefore need less extracted natural uranium to produce electricity. The operation principle of fast breeder reactors is described. They notably directly consume plutonium, allow an easier radioactive waste management as they transform long life radioactive elements into shorter life elements by transmutation. The regeneration process is briefly described, and the various operation modes are evoked (iso-generator, sub-generator, and breeder). Some peculiarities of sodium-cooled reactors are outlined. The Astrid operation principle is described, its main design innovations outlined. Various challenges are discussed regarding safety of supply and waste processing, and the safety of future reactors. Major actors are indicated: CEA, Areva, EDF, SEIV Alcen, Toshiba, Rolls Royce, and Comex. Some key data are indicated: expected lifetime, expected availability rate, cost. The projected site is Marcoule and fast breeder reactors operated or under construction in the world are indicated. The document also proposes an overview of the background and evolution of reactors of 4. generation

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

Energy Technology Data Exchange (ETDEWEB)

You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-07-01

Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

Improvement of research reactor sustainability

International Nuclear Information System (INIS)

Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

2010-01-01

The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

Interfacial rheology of model particles at liquid interfaces and its relation to (bicontinuous) Pickering emulsions

Science.gov (United States)

Thijssen, J. H. J.; Vermant, J.

2018-01-01

Interface-dominated materials are commonly encountered in both science and technology, and typical examples include foams and emulsions. Conventionally stabilised by surfactants, emulsions can also be stabilised by micron-sized particles. These so-called Pickering-Ramsden (PR) emulsions have received substantial interest, as they are model arrested systems, rather ubiquitous in industry and promising templates for advanced materials. The mechanical properties of the particle-laden liquid-liquid interface, probed via interfacial rheology, have been shown to play an important role in the formation and stability of PR emulsions. However, the morphological processes which control the formation of emulsions and foams in mixing devices, such as deformation, break-up, and coalescence, are complex and diverse, making it difficult to identify the precise role of the interfacial rheological properties. Interestingly, the role of interfacial rheology in the stability of bicontinuous PR emulsions (bijels) has been virtually unexplored, even though the phase separation process which leads to the formation of these systems is relatively simple and the interfacial deformation processes can be better conceptualised. Hence, the aims of this topical review are twofold. First, we review the existing literature on the interfacial rheology of particle-laden liquid interfaces in rheometrical flows, focussing mainly on model latex suspensions consisting of polystyrene particles carrying sulfate groups, which have been most extensively studied to date. The goal of this part of the review is to identify the generic features of the rheology of such systems. Secondly, we will discuss the relevance of these results to the formation and stability of PR emulsions and bijels.

Fukushima - calculation of the reactor core inventory and storage pools Dai-ichi 1 to Dai-ichi 4, an estimation of a source term

International Nuclear Information System (INIS)

Krpelanova, M.; Carny, P.

2011-01-01

Inventory of the reactor core and spent fuel storage pool of the reactors at Dai-ichi 1 to Dai-ichi 4 was determined to need a realistic estimate of the source (released into the atmosphere environment) and modelling of radiological impact of the events in Fukushima NPP. Calculations of inventories were carried out by the methodology that is used in systems to support emergency response and crisis management anymore. Calculations were made based on a model that respects knowledge of real fuels and fuel cycles for individual reactors Dai-ichi. Necessary input data for training the model and calculate inventories are obtained from the IAEA PRIS database.

Safety philosophy and safety technology of the Soviet RBMK reactors

International Nuclear Information System (INIS)

Zuend, H.; Jarvis, A.S.; Haennis, H.P.; Tikal, J.

1986-01-01

Safety requirements and control in USSR are outlined. Safety criteria and practical application in the case of the RBMK type reactor Chernobyl-4 are discussed. An overview of the Chernobyl-4 reactor accident including its causes is given. Measures to improve the safety of RBMK reactors are described

Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

International Nuclear Information System (INIS)

Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2007-01-01

Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

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

BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup

International Nuclear Information System (INIS)

1998-01-01

1 - Description of program or function: The system of codes can be used to solve nuclear reactor core static neutronics and reactor history exposure problems. BOLD/VENTURE-4: First order perturbation and time-dependent sensitivity theories can be applied. Control rod positioning may be modeled explicitly and refueling treated with repositioning and recycle. Special capability is coded to model the continuously fueled core and to solve the importance and dominant harmonics problems. The modules of the code system are: VENTNEUT: VENTURE neutronics module; DRIVER and CONTRL: Control module; BURNER: Exposure calculation for reactor core analysis; FILEDTOR: File editor; INPROSER: Input processor; EXPOSURE: BURNER code module; REACRATE: Reaction rate calculation; CNTRODPO: Control rod positioning; FUELMANG: Fuel management positioning and accounting; PERTUBAT: Perturbation reactivity importance analyses; sensitivity analysis; DEPTHMOD: Static and time-dependent perturbation sensitivity analysis. The special processors are: DVENTR: Handles the input to the VENTURE module; DCMACR: Converts CITATION macroscopic cross sections to microscopic cross sections; DCRSPR: Produces input for the CROSPROS module; DUTLIN: Adds or replaces problem input data without exiting the program; DENMAN: Repositions fuel; DMISLY: Miscellaneous tasks. Standard interface files between modules are binary sequential files that follow a standardized format. VENTURE-PC: The microcomputer version is a subset of the mainframe version. The modules and special processors which are not part of VENTURE-PC are: REACRATE, CNTRODPO, PERTUBAT, FUELMANG, DEPTHMOD, DMISLY. 2 - method of solution: BOLD-VENTURE-4: The neutronics problems are solved by applying the multigroup diffusion theory representation of neutron transport applying an over-relaxation inner iteration, outer iteration scheme. Special modeling is used or source correction is done during iteration to solve importance and harmonics problems. No

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

Science.gov (United States)

Cockeram, B. V.; Kammenzind, B. F.

2018-02-01

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

Development of a helical-coil double wall tube steam generator for 4S reactor

International Nuclear Information System (INIS)

Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

2011-01-01

The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

Contribution to the improvement of the evaluation methods of nuclear heating in reactors by using the Monte Carlo code TRIPOLI-4

International Nuclear Information System (INIS)

Peron, Arthur

2014-01-01

Technological irradiation programs carried out in experimental reactors are crucial for the support of the current nuclear fleet in terms of study and anticipation of the behavior under irradiation of fuels and structural materials. These programs make it possible to improve the safety of the current reactors and also to study materials for the new concepts of reactors. Irradiation conditions of materials in experimental reactors must be representative of those of nuclear power plants (NPPs). One of the main advantages of material testing reactors (MTRs) is to be able to carry out instrumented irradiations by adjusting experimental parameters, in particular the neutron flux and the temperature. The control of the parameter temperature of a device irradiated in an experimental reactor requires the knowledge of the nuclear heating (source term) due to the deposition of energy of the photons and the neutrons interacting in the device. A relevant evaluation of this heating is a key data for the thermal studies of design and safety of devices. The objective of this thesis is to improve the methods of the evaluation of nuclear heating in reactors. This work consists of the development of an innovating and complete coupled neutron-photon calculation scheme (allowing to obtain the contribution of neutrons, prompt gamma and decay gamma), mainly based on the 3D, continuous energy TRIPOLI-4 Monte Carlo transport code. An experimental validation of the calculation scheme has been performed, based on calorimetry measurements carried out in the OSIRIS reactor at CEA Saclay. Sensitivity studies have been undertaken to establish the impact of various parameters on nuclear heating calculations (in particular nuclear data) and to fix the final calculation scheme to be closer to the technological irradiation aspects. The thesis work leads to an operational and predictive tool for the nuclear heating estimation, meeting the experimentation needs of research reactors and can be

Corrosion of reactor materials

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-01-15

Much operational experience and many experimental results have accumulated in recent years regarding corrosion of reactor materials, particularly since the 1958 Geneva Conference on the Peaceful Uses of Atomic Energy, where these problems were also discussed. It was, felt that a survey and critical appraisal of the results obtained during this period had become necessary and, in response to this need, IAEA organized a Conference on the Corrosion of Reactor Materials at Salzburg, Austria (4-9 June 1962). It covered many of the theoretical, experimental and engineering problems relating to the corrosion phenomena which occur in nuclear reactors as well as in the adjacent circuits

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Magnetite nanoparticles enhance the performance of a combined bioelectrode-UASB reactor for reductive transformation of 2,4-dichloronitrobenzene.

Science.gov (United States)

Wang, Caiqin; Ye, Lu; Jin, Jie; Chen, Hui; Xu, Xiangyang; Zhu, Liang

2017-09-04

Direct interspecies electron transfer (DIET) among the cometabolism microbes plays a key role in the anaerobic degradation of persistent organic pollutants and stability of anaerobic bioreactor. In this study, the COD removal efficiency increased to 99.0% during the start-up stage in the combined bioelectrode-UASB system (R1) with magnetite nanoparticles addition, which was higher than those in the coupled bioelectrode-UASB (R2; 83.2%) and regular UASB (R3; 71.0%). During the stable stage, the increase of 2,4-dichloronitrobenzene (2,4-DClNB) concentration from 25 mg L -1 to 200 mg L -1 did not affect the COD removal efficiencies in R1 and R2, whereas the performance of R3 was deteriorated obviously. Further intermediates analysis indicated that magnetite nanoparticles enhanced the reductive dechlorination of 2,4-DClNB. High-throughput sequencing results showed that the functional microbes like Syntrophobacter and Syntrophomonas which have been reported to favor the DIET, were predominant on the cathode surface of R1 reactor. It is speculated that the addition of magnetite nanoparticles favors the cooperative metabolism of dechlorinating microbes and electricigens during 2,4-DClNB degradation process in the combined bioelectrode-UASB reactor. This study may provide a new strategy to improve the performance of microbial electrolysis cells and enhance the pollutant removal efficiency.

Development of an innovative reflector drive mechanism using magnetic repulsion force for 4S reactor

International Nuclear Information System (INIS)

Tsuji, K.; Watanabe, M.; Inagaki, H.; Nishikawa, A.; Takahashi, H.; Wakamatsu, M.; Matsumiya, H.; Nishiguchi, Y.

2001-01-01

A small sized fast reactor 4S: (Super Safe Small and Simple) which has a core of 10 - 30 years life time is controlled by reflectors. The reflector is required to be risen at very low speed to make up for the reactivity swing during operation. This report shows the development of an innovative reflector drive mechanism using magnetic repulsion force that can move at a several micrometer per one step. This drive mechanism has a passive shut down capability, and can eliminate reflector drive line. (author)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

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