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Sample records for north anna-1 reactor

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
Seismic design margin evaluation of systems and equipment required for safe shutdown of North Anna, Units 1 and 2, following an SSE (safe-shutdown earthquake) event. Technical report

International Nuclear Information System (INIS)

Desai, K.D.

1981-06-01

The Advisory Committee on Reactor Safeguards recommended that the NRC staff review in detail the capability and available seismic design margin of fluid systems and equipment used in North Anna, Units 1 and 2 to achieve safe shutdown following a site-design safe-shutdown earthquake (SSE). The staff conducted a series of plant visits and meetings with the licensee to view and discuss the seismic design methodology used for systems, equipment and their supports. The report is a description and evaluation of the seismic design criteria, design conservatisms and seismic design margin for North Anna, Units 1 and 2
Corrosion of Zircaloy-clad fuel rods in high-temperature PWRs: Measurement of waterside corrosion in North Anna Unit 1

International Nuclear Information System (INIS)

Balfour, M.G.; Kilp, G.R.; Comstock, R.J.; McAtee, K.R.; Thornburg, D.R.

1992-03-01

Twenty-four peripheral rods and two interior rods from North Anna Unit 1, End-of-Cycle 7, were measured at poolside for waterside corrosion on four-cycle Region 6 assemblies F35 and F66, with rod average burnups of 60 GWD/MTU. Similar measurements were obtained on 24 two-cycle fuel rods from Region 8A assemblies H02 and H10 with average burnups of about 40 GWD/MTU. The Region 6 peripheral rods had been corrosion measured previously after three cycles, at 45 GWD/MTU average burnup. The four-cycle Region 6 fuel rods showed high corrosion, compared to only intermediate corrosion level after three cycles. The accelerated corrosion rate in the fourth cycle was accompanied by extensive laminar cracking and spalling of the oxide film in the thickest regions. The peak corrosion of the two-cycle region 8A rods was 32 μm to 53 μm, with some isolated incipient oxide spalling. In conjunction with the in-reactor corrosion measurements, extensive characterization tests plus long-term autoclave corrosion tests were performed on archive samples of the three major tubing lots represented in the North Anna measurements. The autoclave tests generally showed the same ordering of corrosion by tubing lot as in the reactor; the chief difference between the archive tubing samples was a lower tin content (1.38 percent) for the lot with the lowest corrosion rate compared with a higher tin content (1.58) for the lot with the highest corrosion rate. There was no indication in the autoclave tests of an accelerated rate of corrosion as observed in the reactor
Non-destructive and destructive examination of the retired North Anna 2 Reactor Pressure Vessel Head

International Nuclear Information System (INIS)

Ahluwalia, Kawaljit; Barnes, Robert; Rao, Gutti; Cattant, Francois; Peat, Noel

2006-09-01

Stress corrosion cracking of Alloy 600 and nickel-based weld materials has been the single biggest challenge facing the PWR industry. A fundamental and thorough knowledge was needed to properly explain this phenomenon and develop appropriate mitigation strategies. Non Destructive Examination (NDE) of the North Anna Unit 2 Reactor Vessel Head (RVH) during the 2002 fall outage identified widespread crack indications in the Alloy 600 CRDM penetrations and associated Alloy 182 and 82 J-groove attachment welds. When the Utility decided to replace the RVH, a rare opportunity was provided to the industry to undertake in-depth studies of representative defective CRDM penetrations from a retired RVH. Accordingly, the Materials Reliability Program, undertook a two-phase program on the retired North Anna 2 Alloy 600 RVH. The first phase involved selection and removal of six penetrations from the RVH and penetration decontamination, replication and laboratory NDE. The second phase consisted of a detailed destructive examination of penetration number 54. This paper provides a summary of work undertaken during this program. Criteria for selection of penetrations for removal and procedures used in removal of the penetrations are described. Extreme care was undertaken in decontamination of the penetrations to facilitate laboratory NDE. Penetration number 54 was then subjected to destructive examination to establish a correlation between NDE findings (from both field and laboratory inspections) and actual flaws. Additional objectives of the destructive examination included mechanistic assessment of defect formation and investigation of the annulus environment and wastage characterization. Data obtained from these studies is invaluable in validating safety assessment statements by developing the correlation between field NDE and actual defects. In addition, information gathered from non-destructive and destructive examinations is used to assess accuracy of the NDE techniques
A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

International Nuclear Information System (INIS)

Polley, M.V.; Evans, H.E.

1993-08-01

High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion
North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

International Nuclear Information System (INIS)

Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

1993-01-01

The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress
Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

International Nuclear Information System (INIS)

Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1992-10-01

In a study sponsored by the US Nuclear regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. North Anna was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the North Anna plant
Characterization of solid radwaste generated at Virginia Power's Surry and North Anna stations

International Nuclear Information System (INIS)

Lippard, D.W.; Elguindy, H.; Nelson, R.A.

1987-01-01

This paper describes the results of a study on characterization of the major constituents of the solid radwaste generated at Virginia Power's North Anna and Surry Power Stations. The characterization consisted of identifying the individual constituents of the dry active waste (DAW) and estimating the fraction of the total DAW (both weight and volume) made up by each constituent. The analysis of the characterization of solid waste streams generated at both North Anna and Surry stations has lead to recommending techniques for both source minimization and waste volume reduction
Restart Testing Program for piping following steam generator replacement at North Anna Unit 1

International Nuclear Information System (INIS)

Bain, R.A.; Bayer, R.K.

1993-01-01

In order to provide assurance that the effects of performing steam generator replacement (SGR) at North Anna unit 1 had no adverse impact on plant piping systems, a cold functional verification restart testing program was developed. This restart testing program was implemented in lieu of a hot functional testing program normally used during the initial startup of a nuclear plant. A review of North Anna plant-specific and generic U.S. Nuclear Regulatory Commission requirements for restart testing was performed to ensure that no mandatory hot functional testing was required. This was determined to be the case, and the development of a cold functional test program was initiated. The cold functional test had inherent advantages as compared to the hot functional testing, while still providing assurance of piping system adequacy. The advantages of the cold verification program included reducing risk to personnel from hot piping, increasing the accuracy of measurements with the improvement in work conditions, eliminating engineering activities during the heatup process, and being able to record measurements as construction work was completed allowing for rework or repair of components if required. To ensure the effectiveness of the cold verification program, a project procedure was generated to identify the personnel, equipment, and measurement requirements. An engineering calculation was issued to document the scope of the restart test program, and an additional calculation was developed to provide acceptance criteria for the critical commodity measurements
Final environmental statement related to the operation of North Anna Power Station, Units 1 and 2: (Docket Nos. 50-338 and 50-339)

International Nuclear Information System (INIS)

1980-08-01

The proposed action is the issuance of Operating Licenses to the Virginia Electric and Power Company for the startup and operation of the North Anna Power Station, Units No. 1 and 2, located on Lake Anna in Louisa County, 40 miles east of Charlottesville, Virginia. The information in this second addendum responds to the Commission's directive that the staff address in narrative form the environmental dose commitments and health effects from fuel cycle releases, fuel cycle socioeconomic impacts, and possible cumulative impacts pending further treatment by rulemaking
Use of mock-up training to reduce personnel exposure at the North Anna Unit 1 Steam Generator Replacement Project

Energy Technology Data Exchange (ETDEWEB)

Henry, H.G. [Virginia Power, Mineral, VA (United States); Reilly, B.P. [Bechtel Power Corp., Gaithersburg, MD (United States)

1995-03-01

The North Anna Power Station is located on the southern shore of Lake Anna in Louisa County, approximately forty miles northwest of Richmond, Virginia. The two 910 Mw nuclear units located on this site are owned by Virginia Electric and Power Company (Virginia Power) and Old Dominion Electric Cooperative and operated by Virginia Power. Fuel was loaded into Unit 1 in December 1977, and it began commercial operation in June 1978. Fuel was loaded into Unit 2 in April 1980 and began commercial operation in December 1980. Each nuclear unit includes a three-coolant-loop pressurized light water reactor nuclear steam supply system that was furnished by Westinghouse Electric Corporation. Included within each system were three Westinghouse Model 51 steam generators with alloy 600, mill-annealed tubing material. Over the years of operation of Unit 1, various corrosion-related phenomena had occurred that affected the steam generators tubing and degraded their ability to fulfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators tubing and degraded their ability to fullfill their heat transfer function. Advanced inspection and repair techniques helped extend the useful life of the steam generators, but projections based on the results of the inspections indicated that the existing steam generators would not last their design life and must be repaired. To this end Virginia Power determined that a steam generator replacement (SGR) program was necessary to remove the old steam generator tube bundles and lower shell sections, including the channel heads (collectively called the lower assemblies), and replace them with new lower assemblies incorporating design features that will prevent the degradation problems that the old steam generators had experienced.
Improved waste water polishing in the North Anna clarifier using Durasil ion exchange media

International Nuclear Information System (INIS)

Hensley, M.C.; Sutter, H.G.

1986-01-01

On December 26, 1985, forty-one cubic feet of ion exchange media provided by Duratek was loaded into the North Anna clarifier demineralizer. As of the writing of this manuscript (February 6, 1986), processing using this single loading has continued for 42 straight days without breakthrough. To date, some 5.89 million gallons have been processed at an average flow rate of 140,200 gal/day. Throughputs are 143,700 gal/cu.ft. and still climbing. These figures represent improvements of a factor 20 over the run times experienced with organic resins immediately prior to the introduction of the new materials. This paper describes the testing program of Durasil ion exchange media carried out at North Anna by Duratek which lead to this dramatic improvement in performance
Conformance to Generic Letter 83-28, Items 3.1.3 and 3.2.3, Beaver Valley, Unit No. 1, North Anna, Unit Nos. 1 and 2, Surry Unit Nos. 1 and 2 (Docket Nos. 50-334, 50-338, 50-339, 50-280, and 50-281)

International Nuclear Information System (INIS)

Haroldsen, R.

1985-09-01

This report provides a review of the submittals for Beaver Valley, Unit No. 1, North Anna, Unit Nos. 1 and 2, and Surry, Unit Nos. 1 and 2, for conformance to Generic Letter 83-28, Items 3.1.3 and 3.2.3. The specific plants selected were reviewed as a group because of similarity in type and applicability of review items
Anna Mani

Indian Academy of Sciences (India)

Srimath

National Congress adopted complete independence as its goal, Anna Mani became increasingly drawn to nationalist ... college, Anna Mani obtained a scholarship to do research in physics at the Indian Institute of. Science. ... Anna Mani returned to Independent India in 1948, and she joined the Indian Meteorological.
Final Assessment of Manual Ultrasonic Examinations Applied to Detect Flaws in Primary System Dissimilar Metal Welds at North Anna Power Station

Energy Technology Data Exchange (ETDEWEB)

Anderson, Michael T.; Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Prowant, Matthew S.; Doctor, Steven R.

2014-03-24

PNNL conducted a technical assessment of the NDE issues and protocols that led to missed detections of several axially oriented flaws in a steam generator primary inlet dissimilar metal weld at North Anna Power Station, Unit 1 (NAPS-1). This particular component design exhibits a significant outside-diameter (OD) taper that is not included as a blind performance demonstration mock-up within the industry’s Performance Demonstration Initiative, administered by EPRI. For this reason, the licensee engaged EPRI to assist in the development of a technical justification to support the basis for a site-specific qualification. The service-induced flaws at NAPS-1 were eventually detected as a result of OD surface machining in preparation for a full structural weld overlay. The machining operation uncovered the existence of two through-wall flaws, based on the observance of primary water leaking from the dissimilar metal weld. A total of five axially oriented flaws were detected in varied locations around the weld circumference. The field volumetric examination that was conducted at NAPS-1 was a non-encoded, real-time manual ultrasonic examination. PNNL conducted both an initial assessment, and subsequently, a more rigorous technical evaluation (reported here), which has identified an array of NDE issues that may have led to the subject missed detections. These evaluations were performed through technical reviews and discussions with NRC staff, EPRI NDE Center personnel, industry and ISI vendor personnel, and ultrasonic transducer manufacturers, and laboratory tests, to better understand the underlying issues at North Anna.
Solid radwaste characterization at surry and North Anna Power Stations

International Nuclear Information System (INIS)

Lippard, D.W.

1987-01-01

This paper describes a characterization of the solid radwaste generated at Virginia Power's North Anna and Surry power stations. The primary focus of this characterization was dry active waste (DAW). The characterization, covering the 21-month period from January 1985 through September 1986, was based on information in the station's health physics procurement records, radwaste shipping records, and from interviews with station personnel. The procurement records were the principal source of information for DAW. They were reviewed to determine the quantities of various materials, purchased during the study period, that were expected to become DAW. This provides an upper limit on the quantity in the waste for several major DAW components and a basis for the total amount of other components in the waste. The approach to characterizing DAW discussed in this paper could be implemented and regularly updated by utilizing a computerized procurement records system. If a use code (i.e., contaminated or noncontaminated) is associated with each stock requisition, a characterization could be performed by a computer run. This approach would help track minimization effort effectiveness and would refine the characterization of DAW considerably
Evaluation of Manual Ultrasonic Examinations Applied to Detect Flaws in Primary System Dissimilar Metal Welds at North Anna Power Station

International Nuclear Information System (INIS)

Anderson, Michael T.; Diaz, Aaron A.; Doctor, Steven R.

2012-01-01

During a recent inservice inspection (ISI) of a dissimilar metal weld (DMW) in an inlet (hot leg) steam generator nozzle at North Anna Power Station Unit 1, several axially oriented flaws went undetected by the licensee's manual ultrasonic testing (UT) technique. The flaws were subsequently detected as a result of outside diameter (OD) surface machining in preparation for a full structural weld overlay. The machining operation uncovered the existence of two through-wall flaws, based on the observance of primary water leaking from the DMW. Further ultrasonic tests were then performed, and a total of five axially oriented flaws, classified as primary water stress corrosion cracking (PWSCC), were detected in varied locations around the weld circumference.
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
Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation

Energy Technology Data Exchange (ETDEWEB)

Crawford, Susan L.; Cinson, Anthony D.; MacFarlan, Paul J.; Hanson, Brady D.; Mathews, Royce

2012-08-01

The objective of this investigation was to evaluate the efficacy of ultrasonic testing (UT) for primary water leak path assessments of reactor pressure vessel (RPV) upper head penetrations. Operating reactors have experienced leakage when stress corrosion cracking of nickel-based alloy penetrations allowed primary water into the annulus of the interference fit between the penetration and the low-alloy steel RPV head. In this investigation, UT leak path data were acquired for an Alloy 600 control rod drive mechanism nozzle penetration, referred to as Nozzle 63, which was removed from the North Anna Unit 2 reactor when the RPV head was replaced in 2002. In-service inspection prior to the head replacement indicated that Nozzle 63 had a probable leakage path through the interference fit region. Nozzle 63 was examined using a phased-array UT probe with a 5.0-MHz, eight-element annular array. Immersion data were acquired from the nozzle inner diameter surface. The UT data were interpreted by comparing to responses measured on a mockup penetration with known features. Following acquisition of the UT data, Nozzle 63 was destructively examined to determine if the features identified in the UT examination, including leakage paths and crystalline boric acid deposits, could be visually confirmed. Additional measurements of boric acid deposit thickness and low-alloy steel wastage were made to assess how these factors affect the UT response. The implications of these findings for interpreting UT leak path data are described.
Reclassification of Theileria annae as Babesia vulpes sp. nov.

Science.gov (United States)

Baneth, Gad; Florin-Christensen, Monica; Cardoso, Luís; Schnittger, Leonhard

2015-04-08

Theileria annae is a tick-transmitted small piroplasmid that infects dogs and foxes in North America and Europe. Due to disagreement on its placement in the Theileria or Babesia genera, several synonyms have been used for this parasite, including Babesia Spanish dog isolate, Babesia microti-like, Babesia (Theileria) annae, and Babesia cf. microti. Infections by this parasite cause anemia, thrombocytopenia, and azotemia in dogs but are mostly subclinical in red foxes (Vulpes vulpes). Furthermore, high infection rates have been detected among red fox populations in distant regions strongly suggesting that these canines act as the parasite's natural host. This study aims to reassess and harmonize the phylogenetic placement and binomen of T. annae within the order Piroplasmida. Four molecular phylogenetic trees were constructed using a maximum likelihood algorithm based on DNA alignments of: (i) near-complete 18S rRNA gene sequences (n = 76 and n = 93), (ii) near-complete and incomplete 18S rRNA gene sequences (n = 92), and (iii) tubulin-beta gene sequences (n = 32) from B. microti and B. microti-related parasites including those detected in dogs and foxes. All phylogenetic trees demonstrate that T. annae and its synonyms are not Theileria parasites but are most closely related with B. microti. The phylogenetic tree based on the 18S rRNA gene forms two separate branches with high bootstrap value, of which one branch corresponds to Babesia species infecting rodents, humans, and macaques, while the other corresponds to species exclusively infecting carnivores. Within the carnivore group, T. annae and its synonyms from distant regions segregate into a single clade with a highly significant bootstrap value corroborating their separate species identity. Phylogenetic analysis clearly shows that T. annae and its synonyms do not pertain to Theileria and can be clearly defined as a separate species. Based on the facts that T. annae and its synonyms have not been

Operating US power reactors

International Nuclear Information System (INIS)

Silver, E.G.

1988-01-01

This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees
Anna Karenina, the 'English novel'.Towards the study of anglomania in Leo Tolstoy 's novel Anna Karenina

Directory of Open Access Journals (Sweden)

Natalia Sarana

2014-12-01

Full Text Available The article deals with one storyline of the novel Anna Karenina that stands as the key for the re-search into the significance of Anglomania in the novel. The 1850-1870s in Russian culture is the time of a most intensive formation of the image of the UK as a highly complex combination of real and mythological elements. The novel Anna Karenina, which Tolstoy himself called the novel about modern life, sets forth the fashion for everything ‘English’ in Russian high society in the 1870s with almost documentary precision. The episode the article deals with is Anna Karenina's reading of an English novel. The article looks at different theories of the origin of the novel and suggests a particular novel as the source for the English novel in Anna Karenina. Article argues that the knowledge of the particular English novel contributes not only to the re-search of Anglomania in Anna Karenina and other Tolstoy's works but also gives a significant in-sight into the study of the characters in the novel.
Risk management of the North Anna Power Station Service Water System Preservation Project using the IPE model

International Nuclear Information System (INIS)

Afzali, A.; Donovan, M.D.; Sartain, M.D.; Bankley, A.V.

1993-01-01

This paper discusses the application of the North Anna Power Station Individual Plant Examination (IPE) models in PRA study of the Service Water System Preservation Project (SWSPP). The service water project involves repair and restoration of the Service Water System (SWS) piping and will require excavation of the buried SWS lines in addition to temporarily removing one of the two redundant SWS loops from operation. The SWSPP will be carried out with one or both units in normal operation. The objective of the PRA study was to quantify the risk impact (as measured by the change in Core Damage Frequency (ΔCDF)) of the SWSPP and to identify and evaluate countermeasures to reduce the risk impact of the project activities. The study concluded that the ΔCDF would be acceptable by undertaking preventative measures and by providing additional accident mitigating measures during performance of the SWSPP activities
Anna-Maria Galojan: jah-liikumisega võivad liituda kõik soovijad / Anna-Maria Galojan ; interv. Ishtvan Ban

Index Scriptorium Estoniae

Galojan, Anna-Maria, 1982-

2003-01-01

Ilmunud ka: Severnoje Poberezhje : Subbota 2. aug. lk. 1. Eesti Euroopa liikumise Ida-Virumaa tugikeskuse piirkondlik koordinaator Anna-Maria Galojan oma organisatsioonist, noorte vabatahtlike tööst
Structure of the vortex wake in hovering Anna's hummingbirds (Calypte anna).

Science.gov (United States)

Wolf, M; Ortega-Jimenez, V M; Dudley, R

2013-12-22

Hummingbirds are specialized hoverers for which the vortex wake has been described as a series of single vortex rings shed primarily during the downstroke. Recent findings in bats and birds, as well as in a recent study on Anna's hummingbirds, suggest that each wing may shed a discrete vortex ring, yielding a bilaterally paired wake. Here, we describe the presence of two discrete rings in the wake of hovering Anna's hummingbirds, and also infer force production through a wingbeat with contributions to weight support. Using flow visualization, we found separate vortices at the tip and root of each wing, with 15% stronger circulation at the wingtip than at the root during the downstroke. The upstroke wake is more complex, with near-continuous shedding of vorticity, and circulation of approximately equal magnitude at tip and root. Force estimates suggest that the downstroke contributes 66% of required weight support, whereas the upstroke generates 35%. We also identified a secondary vortex structure yielding 8-26% of weight support. Lift production in Anna's hummingbirds is more evenly distributed between the stroke phases than previously estimated for Rufous hummingbirds, in accordance with the generally symmetric down- and upstrokes that characterize hovering in these birds.
Anna Mani

Indian Academy of Sciences (India)

Srimath

also instilled in her a fierce desire for personal freedom for which she was willing to fight. Thus ... single-authored papers on luminescence of diamonds and ruby. ... It is no surprise that Anna Mani is a success story to which few women (or.
Technical evaluation report on the seven main transformer failures at the North Anna Power Station, Units 1 and 2 (Docket Nos. 50-338, 50-339)

International Nuclear Information System (INIS)

Dalton, K.J.; Kresser, J.V.; Savage, J.W.; Selan, J.C.

1984-01-01

This report documents technical evaluations on various aspects pertaining to the seven main transformer failures at the North Anna Power Station, Units 1 and 2. These reports cover the subjects of Probability Risk Assessment (PRA), Failure Modes and Effects Analysis (FMEA), Root Causes, Protection Systems, Modifications, Failure Statistics, and Generic Aspects. The PRA determined that the contribution from a main transformer failure affecting plant safety systems so as to increase the risk to the public health and safety is negligible. The FMEA determined that a main transformer failure can have primary and secondary effects on plant safety system operation. The evaluation of the Root Causes found that no single common cause contributed to the seven failures. Each failure was found to have specific circumstances for initiating the failure. Both the generator and transformer primary protection systems were found to perform correctly and were designed within industry standards and practices. The proposed modifications resulting from the analyses of the failures will improve system reliability and integrity, and will reduce potentially damaging effects. The failure statistic survey found very limited data bases from which a meaningful correlation could be ascertained. The statistical comparison found no appreciable anomalies with the NAPS failures. The evaluation of all the available information and the results of the separate reports on the main transformer failures found that several generic concerns exist
Revista Annaes de Enfermagem: publicações de enfermeiras sobre pediatria (1932-1941 Revista Annaes de Enfermagem: publicaciones de enfermeras sobre pediatria (1932-1941 Revista Annaes de Enfermagem: nurses' publications about pediatrics (1932-1941

Directory of Open Access Journals (Sweden)

Aline Silva Fontes

2009-02-01

Full Text Available Estudo histórico-social que tem como objeto a produção intelectual de enfermeiras e estudantes de enfermagem relativas à enfermagem pediátrica na revista Annaes de Enfermagem, no período entre 1932-1941. A fonte primária refere-se aos números da revista Annaes de Enfermagem pertencentes ao recorte temporal do estudo, além de relatórios e correspondências. As fontes secundárias estão constituídas de livros, artigos, dissertações e teses relativas à história do Brasil e da enfermagem. A análise dos dados teve o apoio das fontes secundárias e do pensamento do sociólogo francês Pierre Bourdieu. Os resultados evidenciam que a revista Annaes de Enfermagem publicou e divulgou temas importantes sobre a assistência de enfermagem à criança e contribuiu para a visibilidade da enfermeira brasileira junto à comunidade científica.Estudio de naturaleza historico-social que tiene como objeto la producción intelectual de las enfermeras y estudiantes de enfermería relativos a la enfermería pediátrica en el periodico Annaes de Enfermagem, en el período 1932-1941. Las fuentes primarias se refieren a los números de la Revista Annaes de Enfermagem pertenecientes al recorte temporal del estudio, además de los informes y correspondencia. Las fuentes secundarias están compuestas de libros, artículos, tesis y disertaciones sobre la historia de la enfermería de Brasil. El análisis de datos cuenta con el apoyo de fuentes secundarias y el pensamiento del sociólogo francés Pierre Bourdieu. Los resultados muestran que el periodico Annaes de Enfermería publicó sobre las importantes cuestiones de la asistencia de enfermería para el niño y ha contribuydo con la visibilidad de la enfermera brasileña en la comunidad científica.Historic-social study whose object was the intellectual production of nurses and students about pediatric nursing in the journal Annaes de Enfermagem, in the period 1932-1941. The primary source refers to the
Molecular detection of Theileria annae and Hepatozoon canis in foxes (Vulpes vulpes) in Croatia.

Science.gov (United States)

Dezdek, Danko; Vojta, Lea; Curković, Snjezana; Lipej, Zoran; Mihaljević, Zeljko; Cvetnić, Zeljko; Beck, Relja

2010-09-20

An epizootiological field study on tick-borne protozoan infections in foxes (Vulpes vulpes) was carried out in different parts of Croatia. Spleen samples of 191 carcasses of red foxes killed in sanitary hunting, were examined for the presence of hematozoa by polymerase chain reaction (PCR) and subsequent sequencing. The investigation revealed four species of hematozoa in 57 foxes (30%), namely Theileria annae, Theileria sp. 3182/05 and Hepatozoon canis. T. annae was found in 10 foxes (5%), Theileria sp. 3182/05 in a single animal (1%), H. canis in 44 (23%) and Hepatozoon sp. was detected in two foxes (1%). T. annae and H. canis were distributed through all the studied regions, while Theileria sp. 3182/05 and Hepatozoon sp. were restricted to the Zagreb and Zagorje, and Istria regions, respectively. Detection of T. annae in all regions of Croatia indicates the presence of the natural cycle of the parasite and raises the possibility of other vectors other than the proposed Ixodes hexagonus.
BNL ALARA Center: ALARA Notes, No. 9

International Nuclear Information System (INIS)

Khan, T.A.; Xie, J.W.; Beckman, M.C.

1994-02-01

This issue of the Brookhaven National Laboratory's Alara Notes includes the agenda for the Third International Workshop on ALARA and specific instructions on the use of the on-line fax-on-demand service provided by BNL. Other topics included in this issue are: (1) A discussion of low-level discharges from Canadian nuclear plants, (2) Safety issues at French nuclear plants, (3) Acoustic emission as a means of leak detection, (4) Replacement of steam generators at Doel-3, Beaznau, and North Anna-1, (5) Remote handling equipment at Bruce, (6) EPRI's low level waste program, (7) Radiation protection during concrete repairs at Savannah River, (8) Reactor vessel stud removal/repair at Comanche Peak-1, (9) Rework of reactor coolant pump motors, (10) Restoration of service water at North Anna-1 and -2, (11) Steam generator tubing problems at Mihama-1, (12) Full system decontamination at Indian Point-2, (13) Chemical decontamination at Browns Ferry-2, and (14) Inspection methodolody in France and Japan
Clinical findings and normative ocular data for free-living Anna's (Calypte anna) and Black-chinned (Archilochus alexandri) Hummingbirds.

Science.gov (United States)

Moore, Bret A; Maggs, David J; Kim, Soohyun; Motta, Monica J; Bandivadekar, Ruta; Tell, Lisa A; Murphy, Christopher J

2018-02-20

To estimate the prevalence of ocular disease and obtain normative ocular data for free-living hummingbirds. Two hundred and sixty-three free-living, adult Hummingbirds from coastal and inland central California were studied, including Anna's (Calypte anna, n = 186) and Black-chinned (Archilochus alexandri; n = 77) hummingbirds. Slit lamp biomicroscopy and indirect ophthalmoscopy were performed on all individuals. Rebound tonometry, measurement of horizontal palpebral fissure length, and streak retinoscopy were performed on select individuals. Five conscious Anna's Hummingbirds underwent ocular imaging including fundus photography, digital slit lamp photography, and anterior segment and retinal optical coherence tomography. The prevalence of ocular disease in this population was 2.28%. Ocular imaging revealed a thin cornea, shallow anterior chamber, large lens, and a single central, deep convexiclivate fovea. Mean ± SD intraocular pressure was 11.21 ± 2.23 mm Hg. Mean ± SD eyelid length was 2.59 ± 0.19 mm. All eyes were emmetropic or mildly hyperopic with a mean (range) ± SD refractive error of +0.32 (-0.25 to +1) ± 0.33 diopters. Consistent with previous reports, these data suggest that hummingbirds have visual characteristics found in predatory and prey species, as well as a low prevalence of spontaneous ocular disease. This work provides a set of reference values and clinical findings that can be used in the future research on hummingbird vision and ocular disease. It also provides representative diagnostic images of normal birds and demonstrates that advanced ocular imaging can be performed on manually restrained hummingbirds without pharmacologic dilation. © 2018 American College of Veterinary Ophthalmologists.
Hovering and forward flight energetics in Anna's and Allen's hummingbirds.

Science.gov (United States)

Clark, Christopher James; Dudley, Robert

2010-01-01

Aerodynamic theory predicts that the mechanical costs of flight are lowest at intermediate flight speeds; metabolic costs of flight should trend similarly if muscle efficiency is constant. We measured metabolic rates for nine Anna's hummingbirds (Calypte anna) and two male Allen's hummingbirds (Selasphorus sasin) feeding during flight from a free-standing mask over a range of airspeeds. Ten of 11 birds exhibited higher metabolic costs during hovering than during flight at intermediate airspeeds, whereas one individual exhibited comparable costs at hovering and during forward flight up to speeds of approximately 7 m s(-1). Flight costs of all hummingbirds increased at higher airspeeds. Relative to Anna's hummingbirds, Allen's hummingbirds exhibited deeper minima in the power curve, possibly due to higher wing loadings and greater associated costs of induced drag. Although feeding at a mask in an airstream may reduce body drag and, thus, the contributions of parasite power to overall metabolic expenditure, these results suggest that hummingbird power curves are characterized by energetic minima at intermediate speeds relative to hovering costs.
Mani, Miss Anna Modayil

Indian Academy of Sciences (India)

Home; Fellowship. Fellow Profile. Elected: 1960 Section: Earth & Planetary Sciences. Mani, Miss Anna Modayil A.I.I.Sc., FNA 1971-79; Secretary 1977-79. Date of birth: 23 August 1918. Date of death: 16 August 2001. Specialization: Atmospheric Physics and Instrumentation Last known address: c/o Mr K.T. Chandy, 14, ...
Western armament and tactics in the writings of Anna Komnene

Directory of Open Access Journals (Sweden)

Drašković Marko

2006-01-01

Full Text Available In this work, first we reconstructed and commented the western horseman's armament witch Anna Komnene had known (long spear, cross-bow, chain mail "Norman" shield, solarets. Afterwards, we established that Anne knew four types of western horseman's attack (attack in full gallop, attack from back slow march, attack from flank and three types of their battle formation (strewn formation, congested formation, formation of two columns. Also, we commented Anna's knowledge of western siege engines (battering-ram, tortoise catapult, siege tower; we established that Anne knew five types of western siege tower. In the end, we commented several fragments witch show Anna Komnene's knowledge of the western siege tactics.
[Revista Annaes de Enfermagem: nurses' publications about pediatrics (1932-1941)].

Science.gov (United States)

Fontes, Aline Silva; Santos, Tânia Cristina Franco; Oliveira, Alexandre Barbosa de

2009-01-01

Historic-social study whose object was the intellectual production of nurses and students about pediatric nursing in the journal Annaes de Enfermagem, in the period 1932-1941. The primary source refers to the issues of the journal Annaes de Enfermagem considering the established time limits for the study, as well as reports and correspondences. The secondary sources are constituted by books, articles, dissertations and thesis on the Brazilian history of nursing. On data analysis it was used the thought of the French sociologist Pierre Bourdieu. The results show that the journal Annaes de Enfermagem published important issues of nursing care to the child and contributed to the visibility of Brazilian nurses among the scientific community.
Kerjav kunstnik / Anna Hints, Maria Rõhu

Index Scriptorium Estoniae

Hints, Anna, 1982-

2011-01-01

Tallinna Linnagaleriis oli 17.03.-10.04.2011 avatud Anna Hintsi ja Maria Rõhu näitus "Nälg". Autorid näituse taotlusest, raha taotlemisest, aktsioonist "Kerjav kunstnik", külastajate suhtumisest sellesse. Näituse aruanne
76 FR 23846 - Virginia Electric Power Company, LLC, North Anna Power Station, Unit No. 1; Exemption

Science.gov (United States)

2011-04-28

... Commission) now or hereafter in effect. The facility consists of a pressurized-water reactor located in... from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil... prevention, detection, control, extinguishment and preservation of safe shutdown capability is addressed for...
Anna-Maria Galojan - värske hääl poliitikas / Anna-Maria Galojan ; interv. Toomas Verrev

Index Scriptorium Estoniae

Galojan, Anna-Maria, 1982-

2007-01-01

Transiidifirma Transgroup Invest arendusdirektor, Eesti Välispoliitika Instituudi töötaja Anna-Maria Galojan, kes kandideerib Riigikogu valimistel Reformierakonna nimekirjas, räägib oma haridusteest, karjäärist ja poliitilistest eesmärkidest
Eluloofilmide põud paneb rääkima Anna Chapmanist / Andres Laasik

Index Scriptorium Estoniae

Laasik, Andres, 1960-2016

2010-01-01

Ameerika ajalehede käivitatud diskussioonist eluloofilmist, mille võiks vändata spiooniafääriga kuulsaks saanud Anna Chapmanist (Anna Kuštšenko). USA-s, Suurbritannias, Indias ja Soomes valmivatest eluloofilmidest. Austraalia filmist "Mao viimane tantsija"
Evaluation of Proctophyllodes huitzilopochtlii on feathers from Anna's (Calypte anna) and Black-chinned (Archilochus alexandri) Hummingbirds: Prevalence assessment and imaging analysis using light and tabletop scanning electron microscopy.

Science.gov (United States)

Yamasaki, Youki K; Graves, Emily E; Houston, Robin S; OConnor, Barry M; Kysar, Patricia E; Straub, Mary H; Foley, Janet E; Tell, Lisa A

2018-01-01

Proctophyllodes huitzilopochtlii Atyeo & Braasch 1966 (Acariformes: Astigmata: Proctophyllodidae), a feather mite, was found on feathers collected from five hummingbird species in California. This mite has not been previously documented on feathers from Anna's (Calypte anna [Lesson 1829]) or Black-chinned (Archilochus alexandri [Bourcier & Mulsant 1846]) Hummingbirds. A total of 753 hummingbirds were evaluated for the presence of mites by species (Allen's n = 112; Anna's n = 500; Black-chinned n = 122; Rufous n = 18; Calliope n = 1), sex (males n = 421; females n = 329; 3 unidentified), and age (juvenile n = 199; after-hatch-year n = 549; 5 unidentified). Of these 753 hummingbirds evaluated, mites were present on the rectrices of 40.9% of the birds. Significantly more Anna's Hummingbirds were positive for rectricial mites (59.2%) compared with 8.2% of Black-chinned, 0.9% of Allen's, 5.6% of Rufous Hummingbirds, and 0% for Calliope (p-value hummingbird species, male hummingbirds (44.9%) had a higher prevalence of rectricial mites compared to female hummingbirds (36.2%; p-value = 0.004), while juvenile hummingbirds (46.2%) had a non-significantly higher prevalence compared to after-hatch-year hummingbirds (39.0%; p-value = 0.089). On average, the percentage of the long axis of the rachis occupied by mites for the outer rectrices (R4 and R5) was 19%, compared to 11% for inner rectrices (R1 and R2), a significant difference (p-value = hummingbird-feather mite relationship is not well understood, but the specialized TSEM technique may be especially useful in examining natural positioning and developmental aspects of the mites since it allows in situ feather examination of live mites.

Education and training activities at North Carolina State University's PULSTAR reactor

International Nuclear Information System (INIS)

Mayo, C.W.

1992-01-01

Research reactor utilization has been an integral part of the North Carolina State University's (NCSU's) nuclear engineering program since its inception. The undergraduate curriculum has a strong teaching laboratory component. Graduate classes use the reactor for selected demonstrations, experiments, and projects. The reactor is also used for commercial power reactor operator training programs, neutron radiography, neutron activation analysis (NAA), and sample and tracer activation for industrial short courses and services as part of the university's land grant mission. The PULSTAR reactor is a 1-MW pool-type reactor that uses 4% enriched UO 2 pellet fuel in Zircaloy II cladding. Standard irradiation facilities include wet exposure ports, a graphite thermal column, and a pneumatic transfer system. In the near term, general facility upgrades include the installation of signal isolation and computer data acquisition and display functions to improve the teaching and research interface with the reactor. In the longer term, the authors foresee studies of new core designs and the development of beam experiment design tools. These would be used to study modifications that may be desired at the end of the current core life and to undertake the development of new research instruments
Anna Mani – A Student Remembers

Indian Academy of Sciences (India)

Srimath

Developing an Indian ozonesonde was her pet project, assigned to her by K R Ramanathan, ... Her scientific career began with C V Raman at the Indian Institute of ... Independent India was full of opportunities and Anna Mani seized the ...
CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).
Hovering performance of Anna's hummingbirds (Calypte anna) in ground effect.

Science.gov (United States)

Kim, Erica J; Wolf, Marta; Ortega-Jimenez, Victor Manuel; Cheng, Stanley H; Dudley, Robert

2014-09-06

Aerodynamic performance and energetic savings for flight in ground effect are theoretically maximized during hovering, but have never been directly measured for flying animals. We evaluated flight kinematics, metabolic rates and induced flow velocities for Anna's hummingbirds hovering at heights (relative to wing length R = 5.5 cm) of 0.7R, 0.9R, 1.1R, 1.7R, 2.2R and 8R above a solid surface. Flight at heights less than or equal to 1.1R resulted in significant reductions in the body angle, tail angle, anatomical stroke plane angle, wake-induced velocity, and mechanical and metabolic power expenditures when compared with flight at the control height of 8R. By contrast, stroke plane angle relative to horizontal, wingbeat amplitude and wingbeat frequency were unexpectedly independent of height from ground. Qualitative smoke visualizations suggest that each wing generates a vortex ring during both down- and upstroke. These rings expand upon reaching the ground and present a complex turbulent interaction below the bird's body. Nonetheless, hovering near surfaces results in substantial energetic benefits for hummingbirds, and by inference for all volant taxa that either feed at flowers or otherwise fly close to plant or other surfaces. © 2014 The Author(s) Published by the Royal Society. All rights reserved.
Imperatrix Bulgariae Anna-Neda (1277-c.1346

Directory of Open Access Journals (Sweden)

Gjuzelev Vassil

2013-01-01

Full Text Available Si è fato il tentanivo di presentare dettagliatamente la vita di Anna-Neda, la celebre imperatrice bulgara di origine serba. Si è ribadito sulle sue attività e le relazioni con vari personaggi dell’epoca a partire dal matrimonio con Michail III Šišman Asen, despota di Vidin, fino alla morte di lei, probabilmente avvenuta nel 1346 a Dubrovnik. Parole chiave: Impero bulgaro, Vidin, Tǎrnovo, Regno di Napoli. In this article an effort is made to present in detail the life of Anna-Neda, famous Bulgarian empress of Serbian origine. The accent is put on her activities and relations with various characters from the period after her marriage with Michail III Asen Šišman, despot of Vidin, until her death, which probably occurred in 1346 in Dubrovnik.
[Anna Hamilton (1864-1935), the excellence of nursing.

Science.gov (United States)

Diebolt, Évelyne

2017-12-01

A Frenchwoman, Anna Hamilton (1864-1935), daughter of a Franco-English couple, reads with passion the works of Florence Nightingale and takes an interest in nursing. In order to practice it, she first passes the equivalent of a bachelor’s degree in self-education and registers at the Marseille medical school. She wants to prepare a medical thesis on the nursing staff in the hospitals in Europe and is conducting an investigation throughout Europe. She passed her thesis on June 15, 1900 entitled “Considerations on hospital nurses”. This work is immediately published. That same year, she took up a post at the “Maison de santé protestante” in Bordeaux (MSP), founded in 1863. Without managerial staff, she is forced to recruit them abroad. She publishes a professional journal : “La Garde-Malade hospitalière” (1906-1914). Then the war turned the MSP into a military hospital, but the institution continued to receive local paying patients. She was given permission to call the school of nurses : Florence Nightingale School. Anna Hamilton is working with American women to create a medical and social service in Aisne. A graduate, Antoinette Hervey, then opened a medical-social service in Rouen, which would employ up to 30 visiting nurses. In 1916, the MSP received a donation from the domain of Bagatelle. The board of directors wants to sell it, but Anna Hamilton manages to finance a hospital-school thanks to families bereaved by the war and a subscription announced in the “Journal of Nursing”. Other establishments created by former students of the MSP opened : the School-hospital Ambroise Paré in Lille, a nursing home for nurses in Chambon-sur-Lignon in 1927 (the Edith-Seltzer foundation) and a sanatorium in Briançon. After a busy life, Anna Hamilton died of cancer in 1935 and is buried in Bordeaux.
A Teacher's Guide for "Anna Karenina."

Science.gov (United States)

WGBH-TV, Boston, MA.

In 1870, after the successful publication of "War and Peace," Leo Tolstoy began imagining a story about a high-born society woman, "Anna Karenina," who destroys her life by having an adulterous affair. By presenting his adulteress as a sympathetic character, Tolstoy aimed to expose injustices in such Russian institutions as…
Aiamängud vormi ja värviga / Anna-Liisa Unt ; interv. Victoria Parmas

Index Scriptorium Estoniae

Unt, Anna-Liisa, 1982-

2006-01-01

Eesti Maaülikoolis ja Rootsi Põllumajandusülikoolis maastikuarhitektuuri õppinud Anna-Liisa Unt (sünd. 1982) endast, oma aiakujundustest, tüüpilisest eestlase aiast tänapäeval jm. Anna-Liisa Undi tööde loetelu
Eesti goes St. Etienne / Anna Roomet

Index Scriptorium Estoniae

Roomet, Anna

2006-01-01

Näitus Eesti Kunstiakadeemia tootedisaini osakonna magistrantide viimase kahe aasta projektide paremikust esindas Eestit disainibiennaalil St. Etiennis Prantsusmaal. Näituse panid üles Pent Talvet ja Pavel Sidorenko, näitusel osalejad: Anna-Maria Einla, Kätlin Kangur, Triin Kuusler, Andri Laidre, Veiko Liis, Liina-Kai Raivet, P. Sidorenko, P. Talvet, Ingela Viks, Triin Voss
Dietary protein level affects iridescent coloration in Anna's hummingbirds, Calypte anna.

Science.gov (United States)

Meadows, Melissa G; Roudybush, Thomas E; McGraw, Kevin J

2012-08-15

Many animal displays involve colorful ornamental traits that signal an individual's quality as a mate or rival. Brilliant iridescent ornaments are common, but little is currently known about their production cost and signaling value. One potential cost of colorful ornaments is the acquisition of limited dietary resources that may be involved, directly or indirectly, in their production. Protein, the primary component of bird feathers and of many nanostructural components of iridescent traits, is naturally restricted in hummingbird diets (comprised mostly of sugars), suggesting that iridescent coloration may be especially challenging to produce in these animals. In this study, we experimentally investigated the effect of dietary protein availability during molt on iridescent color expression in male Anna's hummingbirds (Calypte anna). We fed captive birds either a 6% (high) or a 3% (low) protein diet and stimulated molt by plucking half the gorget and crown ornaments on each bird as well as the non-ornamental iridescent green tail feathers. We found that birds receiving more protein grew significantly more colorful crown feathers (higher red chroma and redder hue) than those fed the low-protein diet. Diet did not affect gorget coloration, but regrowth of feathers in captivity affected both gorget and crown coloration. Additionally, birds on the high-protein diet grew yellower (higher hue) green tail feathers than birds on the low-protein diet. These results indicate that iridescent ornamental feathers are sensitive to diet quality and may serve as honest signals of nutrition to mates or rivals. Further, because both ornamental and non-ornamental iridescent coloration were affected by conditions during their growth, iridescent color in these birds appears to be generally condition dependent.
[O rzeczach minionych, scripta rerum historicarum annae rutkowska-plachcińska oblata] / Risto Pullat

Index Scriptorium Estoniae

Pullat, Raimo, 1935-

2011-01-01

Arvustus: O rzeczach minionych, scripta rerum historicarum annae rutkowska-plachcińska oblata. (Pod redakeja Marty Mlynarskiej-Kaletynowej i Jerzego Kruppe. Studia i materaly z historii kultury materialney pod redakeja Jerzego Kruppe). Warszawa, 2006. Sisukast ja mitmekülgsest Anna Rutkowska-Plachcinska juubelikogumikust
Seeing things by Anna Margolin

Directory of Open Access Journals (Sweden)

Francuz Aleksandra

2016-12-01

Full Text Available If we recall the most important words for the poetry, we should begin the story with the stones, lily, hands and blood, which, if you look at them with the philological accuracy, are interrelated: situate themselves clearly on opposite banks of the creative process. They talk about the construction of the building consisting of perceptions about their own strength and weakness, ecstasy and congealing in what for centuries the classics poets have tried to fathom: the harmony and clarity, driven by the hope that in life they should above all stick to be beauty. And in addition, they ought to go further than discipline, practice eye and expand the field of view, should not shun from the gusts of the heart, because the heart grows more powerful due to great ideas, passions. Spacious becomes the vision of writing, it is the weave of contradictions - this is how one of the most important twentieth-century poets, who wrote in Yiddish, Anna Margolin sees it. Margolin’s poetry is a clear return towards neoclassicism, building up topics, theses, allusions, ideas taken from ideologically close to her artists of great individuality, Anna Akhmatova (from whom she took her name, Osip Mandelstam, Rainer Maria Rilke, Ezra Pound.
Official portrait of Astronaut Anna L. Fisher

Science.gov (United States)

1985-01-01

Official portrait of Astronaut Anna L. Fisher. Fisher is posing with her helmet on the table in front of her and the American flag appears over the opposite shoulder (34357); Posing with an empty table in front of her and the American flag behind her (34358).
Between practice and theory: Melanie Klein, Anna Freud and the development of child analysis.

Science.gov (United States)

Donaldson, G

1996-04-01

An examination of the early history of child analysis in the writings of Melanie Klein and Anna Freud reveals how two different and opposing approaches to child analysis arose at the same time. The two methods of child analysis are rooted in a differential emphasis on psychoanalytic theory and practice. The Kleinian method derives from the application of technique while the Anna Freudian method is driven by theory. Furthermore, by holding to the Freudian theory of child development Anna Freud was forced to limit the scope of child analysis, while Klein's application of Freudian practice has led to new discoveries about the development of the infant psyche.
Partei tuhandete ja enda kommiraha eest 462 häält / Anna-Maria Galojan ; interv. Rainer Kerge

Index Scriptorium Estoniae

Galojan, Anna-Maria, 1982-

2007-01-01

Intervjuu parlamendivalimistel osalenud politoloogiamagistrant Anna-Maria Galojaniga tema osalemisest valimistel. Vt. samas: Kes on Anna-Maria Galojan? Küsimustele vastab Reformierakonna kampaaniajuht Arto Aas
Professor dr hab. Anna Maria Bujakiewicz

Directory of Open Access Journals (Sweden)

Anna Kujawa

2015-08-01

Full Text Available The article presents the biography and scientific achievements of Professor Anna Bujakiewicz. After receiving her master’s degree and doctorate in biology and mycology from Adam Mickiewicz University in Poznań, Professor Bujakiewicz continued her exciting research and teaching on mycology at her Alma Mater Posnaniensis for more than 50 years. Her publications in this field include many books, articles, and other scholarly reports.
Evaluation of Proctophyllodes huitzilopochtlii on feathers from Anna's (Calypte anna and Black-chinned (Archilochus alexandri Hummingbirds: Prevalence assessment and imaging analysis using light and tabletop scanning electron microscopy.

Directory of Open Access Journals (Sweden)

Youki K Yamasaki

Full Text Available Proctophyllodes huitzilopochtlii Atyeo & Braasch 1966 (Acariformes: Astigmata: Proctophyllodidae, a feather mite, was found on feathers collected from five hummingbird species in California. This mite has not been previously documented on feathers from Anna's (Calypte anna [Lesson 1829] or Black-chinned (Archilochus alexandri [Bourcier & Mulsant 1846] Hummingbirds. A total of 753 hummingbirds were evaluated for the presence of mites by species (Allen's n = 112; Anna's n = 500; Black-chinned n = 122; Rufous n = 18; Calliope n = 1, sex (males n = 421; females n = 329; 3 unidentified, and age (juvenile n = 199; after-hatch-year n = 549; 5 unidentified. Of these 753 hummingbirds evaluated, mites were present on the rectrices of 40.9% of the birds. Significantly more Anna's Hummingbirds were positive for rectricial mites (59.2% compared with 8.2% of Black-chinned, 0.9% of Allen's, 5.6% of Rufous Hummingbirds, and 0% for Calliope (p-value < 0.0001. Across all hummingbird species, male hummingbirds (44.9% had a higher prevalence of rectricial mites compared to female hummingbirds (36.2%; p-value = 0.004, while juvenile hummingbirds (46.2% had a non-significantly higher prevalence compared to after-hatch-year hummingbirds (39.0%; p-value = 0.089. On average, the percentage of the long axis of the rachis occupied by mites for the outer rectrices (R4 and R5 was 19%, compared to 11% for inner rectrices (R1 and R2, a significant difference (p-value = <0.0001. There was a marginal lack of significance for symmetrical distribution of tail mites with the mean left side percentage of long axis of the rachis occupied by mites being 16% and very close to the mean right side score of 18% (p-value = 0.003. The identification of the feather mite species was based on light microscopic morphometry, and mite distribution on feathers was further evaluated using tabletop scanning electron microscopy (TSEM. The hummingbird-feather mite relationship is not well understood
Anna Louropoulou: 'Peri-implantitis vraagt om eenduidige behandelafspraken'

NARCIS (Netherlands)

Louropoulou, A.

2015-01-01

Het aantal geplaatste implantaten neemt toe, maar in de meeste gevallen ontbreekt een adequate nazorg. De NVvP nam het initiatief tot de ontwikkeling van een richtlijn perio-implantitis en zocht hiervoor de samenwerking met de NVOI. Parodontoloog/promovenda Anna Louropoulou: "De richtlijn is
Anna Dziemianko: User-friendliness of Verb Syntax in Pedagogical ...

African Journals Online (AJOL)

Abstract. Anna Dziemianko. User-friendliness of Verb Syntax in Pedagogical Dictionaries of English. 2006, XII + 229 pp. ISBN-13: 978-3-484-39130-7. ISBN- 10: 3-484-39130-8. Lexicographica. Series Maior 130. Tübingen: Max Niemeyer. Price: €124.
Astronaut Anna Fisher demonstrates sleep restraints on shuttle

Science.gov (United States)

1984-01-01

Astronaut Anna L. Fisher demonstrates the versatility of shuttle sleep restraints to accommodate the preference of crewmembers as she appears to have configured hers in a horizontal hammock mode. Stowage lockers, one of the middeck walls, another sleep restraint, a jury-rigged foot and hand restraint are among other items in the frame.

Interview with Dr Anna Matamala

Directory of Open Access Journals (Sweden)

Lucinea Marcelino Villela

2016-09-01

In this interview, which took place in June 2016, Dr Anna Matamala described some details about her long professional experience in Audiovisual Translation, especially in dubbing from English into Catalan, and we talked about many other things like her interest in lexicography, her point of view on some contemporary topics in Audiovisual Translation Studies: the use of technology, the relation between AVT and Accessibility Studies, AVT and Filmmaking fields, the importance of keeping in touch with other countries and even continents outside Europe, and she also gave some advice to the new generation of Translation students.
Interview with Dr Anna Matamala

Directory of Open Access Journals (Sweden)

Lucinea Marcelino Villela

2016-12-01

Full Text Available In this interview, which took place in June 2016, Dr Anna Matamala described some details about her long professional experience in Audiovisual Translation, especially in dubbing from English into Catalan, and we talked about many other things like her interest in lexicography, her point of view on some contemporary topics in Audiovisual Translation Studies: the use of technology, the relation between AVT and Accessibility Studies, AVT and Filmmaking fields, the importance of keeping in touch with other countries and even continents outside Europe, and she also gave some advice to the new generation of Translation students.
Anna Politkovskaja mälestus Euroopas / Marianne Mikko

Index Scriptorium Estoniae

Mikko, Marianne, 1961-

2007-01-01

Ilmunud ka: Virumaa Teataja, 18. sept. 2007, lk. 11; Põhjarannik, 18. sept. 2007, lk. 4; Severnoje Poberezhje, 18. sept. 2007, lk. 4; Sakala, 21. sept. 2007, lk. 2. Euroopa Parlamendi saadik toetas ideed anda Euroopa Parlamendi pressiruumile Anna Politkovskaja nimi, sellega antakse Venemaale ka teada, mida europarlament arvab Vene võimude tegevusest Tšetšeenias
Infantile sexuality: Its place in the conceptual developments of Anna Freud and Donald W. Winnicott.

Science.gov (United States)

Joyce, Angela

2016-06-01

This essay explores the place of infantile sexuality in the theories of Anna Freud and Donald W Winnicott. Both Anna Freud and D.W. Winnicott incorporated and at the same time changed the classical psychoanalytic account of infantile sexuality and the instinctual drives. Whilst Anna Freud remained closer to her father's original conceptualization, she developed a multidimensional model of development which gave the drives a foundational status whist also maintaining their significance in giving meaning and texture to children's subjective experience. Winnicott also retained much of S. Freud's original theorizing except that in a fundamental way he turned it on its head when considering earliest development. For him the establishment of the self was paramount, and the drives and infantile sexuality merely served to give substance to that self. Copyright © 2016 Institute of Psychoanalysis.
Kõrkjasaar Põlvas = Bulrush Island in Põlva / Anna-Liisa Unt

Index Scriptorium Estoniae

Unt, Anna-Liisa, 1982-

2004-01-01

Põlvas 2003. a. korraldatud ideevõistluste tulemused: keskpargi võistlus: II preemia - OÜ Kaunimad Aiad, III - Mart Mets; paisjärve kujunduse konkurss: II - Anna-Liisa Unt, III - Mart Mets. A.-L. Undi realiseeritud võistlustööst, paisjärve saare valguskujundusest. 4 ill
Latin and knitting. Educational and biographical experiences in «My life» by Anna Franchi

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Lucilla Gigli

2014-01-01

Full Text Available The life and works of Anna Franchi remained undiscovered for a long time despite the writer desired to become part of history and to leave a trace both in the socialist and in the emancipationist movements. In the literary production of Anna Franchi, autobiographical elements and fictions become an inseparable issue. Particularly in her biography La mia vita (My life published for the first time in 1940 the writer looked through her life searching for a deeper meaning. Franchi does not limit to describe events and characters but, through narration, she gives an interpretation of her own experience giving in the same time a meaning to it: from her childhood when she absorbs and adopts the ideals of the Risorgimento and the love for art and literature to the long and extremely painful period of her marriage; from journalistic and artistic activities to the political commitment with the socialist party and the movement for women’s rights. Through her works, Franchi provides a political example of free woman Received: 27/05/2013 / Accepted: 20/06/2013 How to reference this article Gigli, L. (2014. Latino e calza. Educazione ed esperienze biografiche ne «La mia vita» di Anna Franchi. Espacio, Tiempo y Educación, 1(1, pp. 97-113. doi: http://dx.doi.org/10.14516/ete.2014.001.001.005
A new species of Tallaperla (Plecoptera: Peltoperlidae) from North Carolina, U.S.A.

Science.gov (United States)

Kondratieff, B.C.; Kirchner, R.F.; Zuellig, R.E.; Lenat, D.R.

2007-01-01

A new species of Tallaperla, T. maiyae, is described from Wilkes County, North Carolina, U.S.A. from two males. The new species is similar to T. maria and T. anna, but can be distinguished by the combination of a prominent spine-like epiproct and brown coloration.
Anna Freud: the Hampstead War Nurseries and the role of the direct observation of children for psychoanalysis.

Science.gov (United States)

Midgley, Nick

2007-08-01

The psychoanalytic tradition of direct observation of children has a long history, going back to the early 20th century, when psychoanalysis and the emerging field of 'child studies' came into fruitful contact in Freud's Vienna. As a leading figure in the attempted integration of direct observation with the new psychoanalytic knowledge emerging from the consulting room, Anna Freud played a crucial role in the emergence of this field. But her major contribution to the theory and practice of observing children came during the Second World War, when she founded the Hampstead War Nurseries. The author describes in detail this important period of Anna Freud's career, and discusses the impact it had on later work. He explores the theoretical contribution that Anna Freud made in the post-war years to the debate about the place of direct observation in psychoanalysis, and concludes that Anna Freud's 'double approach' (direct observation plus analytic reconstruction) still has a great deal to offer as a method of both psychoanalytic research and education.
Melanie Klein and Anna Freud: the discourse of the early dispute.

Science.gov (United States)

Viner, R

1996-01-01

Divisions in the field of the psychoanalysis of children can be traced to a dispute over the infantile super-ego between the theorists Melanie Klein and Anna Freud beginning in 1927. These divisions are understood within the analytic world as the result of scientific disputation between alternative valid theories. An examination of the language, claims, and epistemology of Klein's and Freud's publications in 1927 that marked the public commencement of the conflict, reveals a personalized discourse in which authority was derived from the allegiance, experience, and personal analytic standing of the contestants as much as from theoretical insight. The structure and rhetoric of the debate suggest that, rather than terminating the dispute, the publications of 1927 served to encourage professionalization in child analysis and establish Anna Freud and Melanie Klein as authoritative alternative theorists.
VAMCIS, a new measuring channel for continuous monitoring of leak rates inside PWR steam generators

International Nuclear Information System (INIS)

Champion, G.; Dubail, A.; Lefevre, F.

1988-01-01

In order to assess the primary to secondary leakage, radioactive isotopes, formed in the primary coolant as a result of fission or neutron capture, are usually monitored in the pressurized water reactor (PWR) secondary coolant. Conventional methods mainly based on the detection of 133 Xe, tritium, and 41 Ar are widely used on French Electricite de France (EdF) PWRs. Some years ago, it appeared necessary to improve both leak rate assessments and steam generator tube rupture (SGTR) detection. A volumetric activity measuring channel inside steam (VAMCIS) has been developed for this purpose. The SGTR that occurred at the North Anna PWR has focused the attention of safety authorities on this new measuring channel. It is planned to implement VAMCIS at North Anna in order to check the leak rate variations more accurately
Peter Heller's a Child Analysis with Anna Freud: the significance of the case for the history of child psychoanalysis.

Science.gov (United States)

Midgley, Nick

2012-02-01

A Child Analysis with Anna Freud, a collection of Anna Freud's detailed case notes of her treatment of the young Peter Heller between 1929 and 1932, was first published in English in 1990. Not only does this work give us direct access to Anna Freud's ways of thinking and working at a crucial period in the early history of child analysis; it is also one of the few records of an adult reflecting in depth on the experience of being in analysis as a child. Yet to date this work has received little attention in the psychoanalytic literature. In an attempt to redress this neglect, the Heller case study is placed in the context of Anna Freud's emerging ideas about child analysis. In particular, its significance in the development of her psychoanalytic thinking is investigated in the light of her 1927 book, The Technique of Child Analysis.
Clinical Holistic Medicine: The Case Story of Anna. II. Patient Diary as a Tool in Treatment

Directory of Open Access Journals (Sweden)

Sören Ventegodt

2006-01-01

Full Text Available In spite of extreme childhood sexual and violent abuse, a 22-year-old young woman, Anna, healed during holistic existential therapy. New and highly confrontational therapeutic tools were developed and used to help this patient (like acceptance through touch and acupressure through the vagina. Her vulva and introitus were scarred from repeated brutal rape, as was the interior of her mouth. During therapy, these scars were gently contacted and the negative emotional contents released. The healing was in accordance with the advanced holistic medical toolbox that uses (1 love, (2 trust, (3 holding, and (4 helping the patient to process and integrate old traumas.The case story clearly revealed the philosophical adjustments that Anna made during treatment in response to the severe childhood abuse. These adjustments are demonstrated by her diary, where sentences contain both the feelings and thoughts of the painful present (the gestalt at the time of the abuse, thus containing the essence of the traumas, making the repression of the painful emotions possible through the change in the patients philosophical perspective. Anna's case gives a unique insight into the process of traumatization (pathogenesis and the process of healing (salutogenesis. At the end of the healing, Anna reconnected her existence to the outer world in a deep existential, suicidal crisis and faced her choice of life or death. She decided to live and, in this process, assumed existential responsibility, which made her able to step out of her mental disease. The advanced holistic toolbox seems to help patients heal even from the worst childhood abuse. In spite of the depth of the existential crisis, holistic existential therapy seems to support existential responsibility well and thus safe for the patients.
Intertekstualiteit en die Bose in Kroniek van Perdepoort (Anna M. Louw

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E. Linde

1986-05-01

Patrick White is also an author of religious literature to whom Anna M. Louw is attracted by her own admission. His novels. The solid Mandala and Riders in the Chariot are studied, and similarities with Kroniek van Perdepoort indicated.
Muscle activation patterns and motor anatomy of Anna's hummingbirds Calypte anna and zebra finches Taeniopygia guttata.

Science.gov (United States)

Donovan, Edward R; Keeney, Brooke K; Kung, Eric; Makan, Sirish; Wild, J Martin; Altshuler, Douglas L

2013-01-01

Flying animals exhibit profound transformations in anatomy, physiology, and neural architecture. Although much is known about adaptations in the avian skeleton and musculature, less is known about neuroanatomy and motor unit integration for bird flight. Hummingbirds are among the most maneuverable and specialized of vertebrate fliers, and two unusual neuromuscular features have been previously reported: (1) the pectoralis major has a unique distribution pattern of motor end plates (MEPs) compared with all other birds and (2) electromyograms (EMGs) from the hummingbird's pectoral muscles, the pectoralis major and the supracoracoideus, show activation bursts composed of one or a few spikes that appear to have a very consistent pattern. Here, we place these findings in a broader context by comparing the MEPs, EMGs, and organization of the spinal motor neuron pools of flight muscles of Anna's hummingbird Calypte anna, zebra finches Taeniopygia guttata, and, for MEPs, several other species. The previously shown MEP pattern of the hummingbird pectoralis major is not shared with its closest taxonomic relative, the swift, and appears to be unique to hummingbirds. MEP arrangements in previously undocumented wing muscles show patterns that differ somewhat from other avian muscles. In the parallel-fibered strap muscles of the shoulder, MEP patterns appear to relate to muscle length, with the smallest muscles having fibers that span the entire muscle. MEP patterns in pennate distal wing muscles were the same regardless of size, with tightly clustered bands in the middle portion of the muscle, not evenly distributed bands over the muscle's entire length. Muscle activations were examined during slow forward flight in both species, during hovering in hummingbirds, and during slow ascents in zebra finches. The EMG bursts of a wing muscle, the pronator superficialis, were highly variable in peak number, size, and distribution across wingbeats for both species. In the pectoralis
Astronaut Anna Fisher practices control of the RMS in a trainer

Science.gov (United States)

1984-01-01

Astronaut Anna Lee Fisher, mission specialist for 51-A, practices control of the remote manipulator system (RMS) at a special trainer at JSC. Dr. Fisher is pictured in the manipulator development facility (MDF) of JSC's Shuttle mockup and integration laboratory.
Anna Groundwater, The Scottish Middle March 1573-1625: Power, Kinship, Allegiance

Directory of Open Access Journals (Sweden)

Janine van Vliet

2011-09-01

Full Text Available Anna Groundwater. The Scottish Middle March 1573-1625: Power, Kinship, Allegiance. Royal Historical Society Studies in History, New Series. Woodbridge, UK: The Boydell Press, 2010. Pp. 236. ISBN 978-0-86193-307-5. £50.00; US$90.00.
Wingbeat kinematics and motor control of yaw turns in Anna's hummingbirds (Calypte anna).

Science.gov (United States)

Altshuler, Douglas L; Quicazán-Rubio, Elsa M; Segre, Paolo S; Middleton, Kevin M

2012-12-01

The biomechanical and neuromuscular mechanisms used by different animals to generate turns in flight are highly variable. Body size and body plan exert some influence, e.g. birds typically roll their body to orient forces generated by the wings whereas insects are capable of turning via left-right wingbeat asymmetries. Turns are also relatively brief and have low repeatability, with almost every wingbeat serving a different function throughout the change in heading. Here we present an analysis of Anna's hummingbirds (Calypte anna) as they fed continuously from an artificial feeder revolving around the outside of the animal. This setup allowed for examination of sustained changes in yaw without requiring any corresponding changes in pitch, roll or body position. Hummingbirds sustained yaw turns by expanding the wing stroke amplitude of the outer wing during the downstroke and by altering the deviation of the wingtip path during both downstroke and upstroke. The latter led to a shift in the inner-outer stroke plane angle during the upstroke and shifts in the elevation of the stroke plane and in the deviation of the wingtip path during both strokes. These features are generally more similar to how insects, as opposed to birds, turn. However, time series analysis also revealed considerable stroke-to-stroke variation. Changes in the stroke amplitude and the wingtip velocity were highly cross-correlated, as were changes in the stroke deviation and the elevation of the stroke plane. As was the case for wingbeat kinematics, electromyogram recordings from pectoral and wing muscles were highly variable, but no correlations were found between these two features of motor control. The high variability of both kinematic and muscle activation features indicates a high level of wingbeat-to-wingbeat adjustments during sustained yaw. The activation timing of the muscles was more repeatable than the activation intensity, which suggests that the former may be constrained by harmonic
Kundnöjdheten i Hotel Anna

OpenAIRE

Boije, Daniel

2014-01-01

I dagens läge värderas kundrelationer högt inom marknadsföringen. Kundnöjdhet förknippas med lönsamhet och ses som en långsiktig investering. Hotel Anna är ett litet hotell i centrala Helsingfors och fungerar som min uppdragsgivare. Syftet med detta arbete är att utreda kundnöjdheten samt hur kundbetjäning och kvalitet upplevs i hotellet. Delsyftet är också att hitta kundnöjdhets förbättringsförslag för hotellet. Den teoretiska referensramen består av teori som behandlar tjänster, kvalitet, k...
Study On In Vitro Variation Generation Of Cymbidium Labell Anna Belle By Gamma Co-60 Irradiation

International Nuclear Information System (INIS)

Nguyen Huynh Phuong Uyen; Vo Thi Thu Ha; Le Quang Luan

2011-01-01

Radiation ionization is one of useful methods for mutation breeding of plants. In this study, γ-ray (Co-60) was used for generating morphological variations of Cymbidium Labell Anna Belle. The lethal doses of 50% samples (LD 50 ) after 4 months cultivation were determined at 35.0 Gy for protocorm like bodies (PLBs), 37.4 Gy for in vitro shoot bubs and 70.1 Gy for plantlets. PLBs were found as suitable samples for generation of variation of mentioned Cymbidium. The dose range from 20 to 50 Gy was determined with a high ability of variation on C. Labell Anna Belle. The highest frequency of variation was obtained ca. 3.82 x 10 -3 at the dose of 30 Gy. 157 in vitro variant plantlets were selected in M 1 V 1 generation, among variations, the variations were mainly found with 5 following types: removal of chlorophyll (8 plantlets), the leaves become shorter (64 plantlets), the leaves become longer (64 plantlets), increase of leaf number (11 plantlets) and the colour of pericardium was changed from green into violet (10 plantlets). Twenty one variant lines were selected in M 1 V 2 generation. (author)
Detection of Babesia annae DNA in lung exudate samples from Red foxes (Vulpes vulpes) in Great Britain.

Science.gov (United States)

Bartley, Paul M; Hamilton, Clare; Wilson, Cari; Innes, Elisabeth A; Katzer, Frank

2016-02-12

This study aimed to determine the prevalence of Babesia species DNA in lung exudate samples collected from red foxes (Vulpes vulpes) from across Great Britain. Babesia are small piroplasmid parasites which are mainly transmitted through the bite of infected ticks of the family Ixodidae. Babesia can cause potentially fatal disease in a wide-range of mammalian species including humans, dogs and cattle, making them of significant economic importance to both the medical and veterinary fields. DNA was extracted from lung exudate samples of 316 foxes. A semi-nested PCR was used to initially screen samples, using universal Babesia-Theileria primers which target the 18S rRNA gene. A selection of positive PCR amplicons were purified and sequenced. Subsequently specific primers were designed to detect Babesia annae and used to screen all 316 DNA samples. Randomly selected positive samples were purified and sequenced (GenBank accession KT580786). Clones spanning a 1717 bp region of the 18S rRNA gene were generated from 2 positive samples, the resultant consensus sequence was submitted to GenBank (KT580785). Sequence KT580785 was used in the phylogenetic analysis Babesia annae DNA was detected in the fox samples, in total 46/316 (14.6%) of samples tested positive for the presence of Babesia annae DNA. The central region of England had the highest prevalence at 36.7%, while no positive samples were found from Wales, though only 12 samples were tested from this region. Male foxes were found to have a higher prevalence of Babesia annae DNA than females in all regions of Britain. Phylogenetic and sequence analysis of the GenBank submissions (Accession numbers KT580785 and KT580786) showed 100% identity to Babesia sp.-'Spanish Dog' (AY534602, EU583387 and AF188001). This is the first time that Babesia annae DNA has been reported in red foxes in Great Britain with positive samples being found across England and Scotland indicating that this parasite is well established within the

[Participation of the Anna Nery School in the Constitutionalist Revolution of 1932].

Science.gov (United States)

de Almeida Filho, Antonio José; Santos, Tânia Cristina Franco

2003-01-01

This is a historical-social research project. The main objective is to present the participation of the Anna Nery Nursing School in the medical assistance positions in the state of Sao Paulo during the Constitutionalist Revolution of 1932. The objective of the present investigation is to describe how the teachers and students of the Anna Nery Nursing School participated in the different operation fronts during this war and to analyse the implications of the performance of nurses and students of this School. Our main documental resource were written and photographical documents that belong to the Centre of Documentation of the EEA/UFRJ. The secondary source were articles and books that about the history of Brazil and Brazilian nursing. This investigation evidenced the importance of the nurse's work during times of crisis and it also made possible for the EEAN to earn symbolic profits.
Vos, Faust, Voss: raakpunte tussen Vos deur Anna M. Louw en enkele ander tekste

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P. John

2005-07-01

Full Text Available Vos, Faust, Voss: Points of contact between Vos by Anna M. Louw and selected other texts This article explores points of contact between “Vos” by Anna M. Louw and a number of related texts, including the following: the book “Job” from the Christian Bible, “Faust” by Goethe, “Voss” by Patrick White and texts forming part of the Gnostic tradition. The analysis describes similarities and differences, arguing that the implication of the points of contact between the various texts is that Vos suggests that the heritage of the Christian Reformation is not adequate for an understanding of life in South Africa, and it has to be supplemented with perspectives from other traditions.
Nuclear Power 2010 Program Dominion Virginia Power Cooperative Project U.S. Department of Energy Cooperative Agreement DE-FC07-05ID14635 Construction and Operating License Demonstration Project Final Report

International Nuclear Information System (INIS)

Grecheck, Eugene S.; Batalo, David P.

2010-01-01

This report serves to summarize the major activities completed as part of Virginia Electric and Power Company's North Anna construction and operating license demonstration project with DOE. Project successes, lessons learned, and suggestions for improvement are discussed. Objectives of the North Anna COL project included preparation and submittal of a COLA to the USNRC incorporating ESBWR technology for a third unit a the North Anna Power Station site, support for the NRC review process and mandatory hearing, obtaining NRC approval of the COLA and issuance of a COL, and development of a business case necessary to support a decision on building a new nuclear power plant at the North Anna site.
Abjeção em textos de André Sant’Anna

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Leandro Salgueirinho

2009-07-01

Full Text Available Trata-se de uma leitura de textos do escritor André Sant’Anna a partir da problematização que, em The Return of the Real (1986, o crítico de arte norte-americano Hal Foster faz da noção de abjeto e da dita abject art.
Flying in the rain: hovering performance of Anna's hummingbirds under varied precipitation.

Science.gov (United States)

Ortega-Jimenez, Victor Manuel; Dudley, Robert

2012-10-07

Flight in rain represents a greater challenge for smaller animals because the relative effects of water loading and drop impact are greater at reduced scales given the increased ratios of surface area to mass. Nevertheless, it is well known that small volant taxa such as hummingbirds can continue foraging even in extreme precipitation. Here, we evaluated the effect of four rain intensities (i.e. zero, light, moderate and heavy) on the hovering performance of Anna's hummingbirds (Calypte anna) under laboratory conditions. Light-to-moderate rain had only a marginal effect on flight kinematics; wingbeat frequency of individuals in moderate rain was reduced by 7 per cent relative to control conditions. By contrast, birds hovering in heavy rain adopted more horizontal body and tail positions, and also increased wingbeat frequency substantially, while reducing stroke amplitude when compared with control conditions. The ratio between peak forces produced by single drops on a wing and on a solid surface suggests that feathers can absorb associated impact forces by up to approximately 50 per cent. Remarkably, hummingbirds hovered well even under heavy precipitation (i.e. 270 mm h(-1)) with no apparent loss of control, although mechanical power output assuming perfect and zero storage of elastic energy was estimated to be about 9 and 57 per cent higher, respectively, compared with normal hovering.
Anna Groundwater, The Scottish Middle March 1573-1625: Power, Kinship, Allegiance

OpenAIRE

Janine van Vliet

2011-01-01

Anna Groundwater. The Scottish Middle March 1573-1625: Power, Kinship, Allegiance. Royal Historical Society Studies in History, New Series. Woodbridge, UK: The Boydell Press, 2010. Pp. 236. ISBN 978-0-86193-307-5. £50.00; US$90.00.
Nuclear Power 2010 Program Dominion Virginia Power Cooperative Project U.S. Department of Energy Cooperative Agreement DE-FC07-05ID14635 Construction and Operating License Demonstration Project Final Report

Energy Technology Data Exchange (ETDEWEB)

Eugene S. Grecheck

2010-11-30

This report serves to summarize the major activities completed as part of Virginia Electric and Power Company's North Anna construction and operating license demonstration project with DOE. Project successes, lessons learned, and suggestions for improvement are discussed. Objectives of the North Anna COL project included preparation and submittal of a COLA to the USNRC incorporating ESBWR technology for a third unit a the North Anna Power Station site, support for the NRC review process and mandatory hearing, obtaining NRC approval of the COLA and issuance of a COL, and development of a business case necessary to support a decision on building a new nuclear power plant at the North Anna site.
Digital instrument for reactivity measurements in a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Chwaszczewski, S [Institute of Nuclear Research, Warsaw (Poland)

1979-07-01

An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given.
NRC Information Notice No. 93-22: Tripping of Klockner-Moeller molded-case circuit breakers due to support lever failure

International Nuclear Information System (INIS)

Grimes, B.K.

1993-01-01

On July 1, 1992, Virginia Electric and Power Company (VEPCO), licensee for North Anna Power Station (North Anna), Units 1 and 2, reported to the NRC that over a three-month period in 1992, three K-M model NZM6-63, 480-Vac MCCBs had tripped without appreciable load or fault condition or other electrical or mechanical transient. The three failed MCCBs were located in the cable vault and tunnel area of North Anna-2 and supplied power to motor operated valves in the charging and safety injection systems. The switch handles were found in the trip-free position and the breakers could not be relatched and reclosed. Examination of the internals of one of the three failed breakers revealed that its support lever (also described by VEPCO as a ''spring arm''), located in the rear compartment of the case, had fractured. This caused the breaker to trip and to become incapable of being reclosed. The NRC has determined that the support levers only in K-M model NZM6, NZM6b, and NZMH6 MCCBs, rated for 100 amperes and below and manufactured from 1972 on, have been made of the same type of plastic used in the support levers of the MCCBs that failed at North Anna-2
Influence of canopy form on growth, flowering, and storage potential of ,anna, apple

International Nuclear Information System (INIS)

Zahran, A.A.

2000-01-01

this study was carried out during the two successive seasons of 1997 and 1998. two years old anna apple trees (budded on M.M. 106 semi - dwarfing root - stocks) planted in El Khattatbah, Monoufiah Governorate were used in this investigation. this study aimed to evaluate the influence of two different tree canopy forms (central leaders and vase) and tipping treatment on the yield and some fruit quality characteristics of anna apple fruits and the effect of irradiation on some of these characteristics during a limited period under cold storage. results revealed improved light conditions in trained trees compared to untrained controls. open vase trained trees produce less fruits which where bigger, softer, with more T.S.S. and sugar content, and less acidity, compared to central leader harvested fruits.it was also found that tipping application resulted in excessive vegetative growth and reduced fruit numbers which were bigger, softer, and contained more T.S.S., and sugars, and less acidity, compared to fruits harvested from untipped trees
Anna Politkovskaya: Rússlands Pútíns

Directory of Open Access Journals (Sweden)

Pétur Berg Matthíasson

2009-12-01

Full Text Available Í umsögn gagnrýnanda kemur meðal annars eftirfarandi fram: Fram kemur í lok bókarinnar að Anna hafi illan bifur á Pútín. Það kemur lesandanum í sjálfu sér ekki á óvart enda bókin upptalning á skelfilegum atburðum sem lýsa hvers konar stjórnvöldum hann fer fyrir, fyrir utan það að Pútín gerir lítið sem ekkert til að breyta því til hins betra. Þessi bók er nauðsynleg fyrir alla þá sem vilja skyggnast bak við tjöldin og sjá Rússland eins og það er í raun og veru.
“Speaking German Like Nobody’s Business”: Anna May Wong, Walter Benjamin, and the Possibilities of Asian American Cosmopolitanism

OpenAIRE

Lim, Shirley Jennifer

2012-01-01

In the summer of 1928 in Berlin, the noted German Jewish philosopher Walter Benjamin (1892–1940) and Chinese American actress Anna May Wong (1905–1961) shared an unlikely encounter that set in relief European and American conceptions of modernity as well as white European intellectual and American racial minority cosmopolitanisms. On July 6, 1928, Benjamin published the results as “Gespräch mit Anne May Wong” [“Speaking with Anna May Wong: A Chinoiserie from the Old West”] on the front page o...
Physical and geometrical parameters of ANNA critical assemblies. Pt. 2

International Nuclear Information System (INIS)

Malewski, S.; Dabrowski, C.

1973-01-01

An extended analysis of four critical configurations of ANNA Assembly has been performed. Diffusion parameters of the thermal group and of one or three epithermal groups have been determined. Using these data the critical calculations have been carried out and the main neutron density distributions presented. The role of some neutron processes in these systems and their influence on integral parameters has been considered. The calculated quantities have been compared with the available experimental data. (author)
A digital instrument for reactivity measurements in a nuclear reactor

International Nuclear Information System (INIS)

Chwaszczewski, S.

1979-01-01

An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given. (author)
Perturbation method for experimental determination of neutron spatial distribution in the reactor cell

International Nuclear Information System (INIS)

Takac, S.M.

1972-01-01

The method is based on perturbation of the reactor cell from a few up to few tens of percent. Measurements were performed for square lattice calls of zero power reactors Anna, NORA and RB, with metal uranium and uranium oxide fuel elements, water, heavy water and graphite moderators. Character and functional dependence of perturbations were obtained from the experimental results. Zero perturbation was determined by extrapolation thus obtaining the real physical neutron flux distribution in the reactor cell. Simple diffusion theory for partial plate cell perturbation was developed for verification of the perturbation method. The results of these calculation proved that introducing the perturbation sample in the fuel results in flattening the thermal neutron density dependent on the amplitude of the applied perturbation. Extrapolation applied for perturbed distributions was found to be justified
Anna Pavlova, danzatrice eurasiana? (I parte

Directory of Open Access Journals (Sweden)

Silvia Mei

2010-12-01

Full Text Available Abstract – IT Prima diva nella costellazione di nuove star dello spettacolo novecentesco, Anna Pavlova visse unicamente nello spirito e nella missione della danza come forma d’arte e stile di vita. Senza temere i rischi del solipsismo, si lancia nell’avventura di una compagnia tutta sua dopo il clamore parigino dei Ballets Russes. Il suo coraggio, la sua curiosità, il suo genio artistico la portano a viaggiare per i cinque continenti senza sosta, a scoprire danze tradizionali, a praticarle essa stessa, come quelle kabuki, apprese non senza sofferenze in Giappone nel 1922. Presentata nella vulgata come paladina della tradizione classicoaccademica, Pavlova può ostentare una biografia degna delle grandi danzatrici moderne a lei contemporanee, come una Duncan o una Ruth St. Denis (con quest’ultima in rapporto di competizione a distanza intorno al soggetto di Radha. Insieme ad una figura controversa e intrigante come quella del danzatore indiano Uday Shankar, crea un trittico di danze di ispirazione orientale all’insegna del rigore filologico e dell’integrazione linguistica, traccia della sua apertura e del rispetto di tradizioni teatrali altre. Abstract – EN Anna Pavlova, the first diva within the stars firmament of the 20th century, spent her life exclusively in the spirit of dance, intended as an art form and a life style, as if it were a mission. She threw herself into the adventure of leading her own company, after the Ballets Russes' European acclaim. Her courage, curiosity and artistic genius took her through five continents, endlessly searching for traditional dances that she learned with tremendous physical suffering, as in Japan with kabuki in 1922. She’s not simply a paladin of the academic tradition. She can in fact claim a biography and dance practice worthy of the most famous modern dancers who lived in her age, such as Duncan and St. Denis (with whom she was in a special competition, though after years, on the subject
CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2018-01-01

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.
Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-04-01

This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.
Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

International Nuclear Information System (INIS)

1997-04-01

This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee's annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs
Anna Lampérière, solidarité et citoyenneté féminine sous la Troisième République

Directory of Open Access Journals (Sweden)

Anne R. Epstein

2008-12-01

Full Text Available Cet article examine les idées de l’institutrice Anna Lampérière, l’une des rares femmes à avoir contribué à la discussion théorique sur l’idéologie de la solidarité pendant la Troisième République. Vue dans le contexte de son parcours personnel et de son engagement civique, la pensée d’A. Lampérière offre un aperçu inédit d’un modèle solidariste de citoyenneté féminine sans droit de vote. Au sein d’une culture d’associations florissante au début du XXe siècle, elle participe à la création et au fonctionnement d’importantes associations mixtes qui, par l’éducation civique, visent la mise en place de l’idéologie solidariste de Léon Bourgeois. Se proclamant anti-féministe, Anna Lampérière se présente néanmoins comme une défenseuse de la cause des femmes et du progrès social et, comme beaucoup de ses contemporains, elle place la réforme scolaire au cœur de son projet républicain. Après une analyse de la vision sociale d’Anna Lampérière, l’article examine les pratiques civiques et la structure mixte selon les sexes de la Société pour l’éducation sociale, une association pour la promotion du solidarisme, dont elle a été Secrétaire générale. L’idéologie solidariste et sa pratique avaient un attrait pour des femmes républicaines comme Anna Lampérière qui avaient une vocation civique mais qui n’étaient pas attirées par un modèle égalitaire de la citoyenneté.Anna Lampérière, solidarité and female citizenship in the Third Republic This article explores the ideas of educator Anna Lampérière, one of the rare women to contribute significantly to the theoretical discussion about the ideology of solidarity during the Third Republic. Viewed in tandem with her own personal history and civic activism, Lampérière’s thought provides unique insight into a possible solidarist model of female citizenship without voting rights. Operating within the flourishing associational

KR’PTA. Samtidspoesin och Derrida : Spår och ärrbildningar hos Johannes Heldén, Ingrid Storholmen och Anna Hallberg

OpenAIRE

Schmidt, Lisa

2006-01-01

Lisa Schmidt, KR’PTA. Samtidspoesin och Derrida. Spår och ärrbildningar hos Johannes Heldén, Ingrid Storholmen och Anna Hallberg. (CR’PT. Contemporary Poetry and Derrida: Traces and Scarring in the Poetry of Johannes Heldén, Ingrid Storholmen and Anna Hallberg.) Through the analyses of three contemporary Nordic poets whose work challenges the boundaries of literature and even the laws of grammar, I draw attention to the term linguistic materialism. I also sketch an historical line between t...
Security and emergency response for a remote distributed reactor fleet in Canada's far north

International Nuclear Information System (INIS)

Harris, J.

2013-01-01

The far north has become an area of great interest to SMR developers in the last several years. Remote and isolated communities with some small exceptions have not been the traditional territory for nuclear reactors. The technology however has now reached the point where their deployment in some of these nontraditional areas can now be considered possible. We do need to ask what would be the challenges of providing security and emergency support in these regions, and how could it be done effectively and cost efficiently? (author)
Kahevahel : Kui hajusad on kultuuriruumi piirid? / Anna Verschik, Tiina Kirss, Mikko Lagerspetz ; intervjueerinud Daniele Monticelli, Merle Karro-Kalberg

Index Scriptorium Estoniae

Verschik, Anna, 1968-

2015-01-01

Kultuurist, kultuuriruumist, kultuuris olemisest ja sealt pärinemisest arutlevad TLÜ professori Daniele Monticelli eestvõttel TLÜ professorid Anna Verschik ja Tiina Kirss ning Turu Åbo Akadeemia professor Mikko Lagerspetz
Paediatrics and psychoanalysis--Miss Anna Freud.

Science.gov (United States)

1983-01-01

Miss Anna Freud died during the winter at the age of 86. She had been a pioneer in the understanding of children through psychoanalysis and a great champion of the rights of children. Her life began in Vienna as the youngest child of Sigmund Freud, and her early work with children was in Austria. In 1938, because of the Nazi régime and even though she was nursing her father during his terminal illness, she had to escape with him to London. Her work with homeless children and with those in residential nurseries in London during the second world war is well known, as is her work on child development and psychopathology in the postwar years. But one less well known aspect of her life that was of immense importance to a few fortunate British paediatricians was the 'paediatric group' that she ran for over a quarter of a century and which Dr Christine Cooper recalled at the memorial meeting in London earlier this year. PMID:6344806
Formation of Barents Sea Branch Water in the north-eastern Barents Sea

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Vidar S. Lien

2013-09-01

Full Text Available The Barents Sea throughflow accounts for approximately half of the Atlantic Water advection to the Arctic Ocean, while the other half flows through Fram Strait. Within the Barents Sea, the Atlantic Water undergoes considerable modifications before entering the Arctic Ocean through the St. Anna Trough. While the inflow area in the south-western Barents Sea is regularly monitored, oceanographic data from the outflow area to the north-east are very scarce. Here, we use conductivity, temperature and depth data from August/September 2008 to describe in detail the water masses present in the downstream area of the Barents Sea, their spatial distribution and transformations. Both Cold Deep Water, formed locally through winter convection and ice-freezing processes, and Atlantic Water, modified mainly through atmospheric cooling, contribute directly to the Barents Sea Branch Water. As a consequence, it consists of a dense core characterized by a temperature and salinity maximum associated with the Atlantic Water, in addition to the colder, less saline and less dense core commonly referred to as the Barents Sea Branch Water core. The denser core likely constitutes a substantial part of the total flow, and it is more saline and considerably denser than the Fram Strait branch as observed within the St. Anna Trough. Despite the recent warming of the Barents Sea, the Barents Sea Branch Water is denser than observed in the 1990s, and the bottom water observed in the St. Anna Trough matches the potential density at 2000 m depth in the Arctic Ocean.
THE IMPACT OF TRANSLATION TECHNIQUES AND SUBTITLING RULES TO THE EQUIVALENCE OF MEANING OF CULINARY TERMS IN THE FRESH WITH ANNA OLSON SEASON 1 SUBTITLE

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Hetty Hartati Novita

2017-12-01

Full Text Available The emergence of cable TV in Indonesia has made this author interested in focusing on the topic of culinary terms translation in the subtitle of Fresh with Anna Olson Season 1 program broadcasted on Asian Food Channel station. In this paper, this author analyzes the translation techniques suggested by Molina and Albir used by the translators to create the more accurate, natural, and communicative subtitle even though there are subtitling rules to follow. The result is, most of the translated terms have equivalence in meaning between the ST and the TT, while the ones which are not equivalent are due to the application of two particular techniques, Discursive Creation and Adaptation.
African American Women Scholars and International Research: Dr. Anna Julia Cooper's Legacy of Study Abroad

Science.gov (United States)

Evans, Stephanie Y.

2009-01-01

EIn this article, the author presents a little-known but detailed history of Black women's tradition of study abroad. Specifically, she situates Dr. Anna Julia Cooper within the landscape of historic African American students who studied in Japan, Germany, Jamaica, England, Italy, Haiti, India, West Africa, and Thailand, in addition to France. The…
The NCSU [North Carolina State Univ.] freon PWR [pressurized water reactor] loop

International Nuclear Information System (INIS)

Caves, J.R.; Doster, J.M.; Miller, G.D.; Wehring, B.W.; Turinsky, P.J.

1989-01-01

The nuclear engineering department at North Carolina State University has designed and constructed an operating scale model of a pressurized water reactor (PWR) nuclear steam supply system (NSSS). This facility will be used for education, training, and research. The loop uses electric heaters to simulate the reactor core and Freon as the primary and secondary coolant. Viewing ports at various locations in the loop allow the students to visualize flow regimes in normal and off-normal operating conditions. The objective of the design effort was to scale the thermal-hydraulic characteristics of a two-loop Westinghouse NSSS. Provisions have been made for the simulation of various abnormal occurrences. The model is instrumented in much the same manner as the actual NSSS. Current research projects using the loop include the development of adaptive expert systems to monitor the performance of the facility, diagnose mechanical faults, and to make recommendations to operators for mitigation of accidents. This involves having thermal-hydraulics and core-physics simulators running faster than real time on a mini-supercomputer, with operating parameters updated by communication with the data acquisition and control computer. Further opportunities for research will be investigated as they arise
Training courses at VR-1 reactor

International Nuclear Information System (INIS)

Sklenka, L.; Kropik, M.

2006-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)
Accounting history and the feminine gender: The case Anna Jansen, Queen of Maranhão (19th Century

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Eliane Silva Sampaio

2017-07-01

By analyzing the manuscripts and newspapers of the time it was concluded that Anna Jansen became the owner of one of the largest fortunes in the region after becoming the widow of her first husband, Coronel Izidoro Pereira, and through this fortune has achieved a prominent position in the Maranhão society. Her performance represents a breaking of paradigms that tend to cancel the participation of women in economic and political activities, and led to her recognition as a woman far ahead of the time in which she lived. Although it was not possible to find technical records kept by Anna Jansen throughout her life, it has been possible through primary documents to gain knowledge that she had great dexterity for the administration of her properties and assets.
Current status of the Thai Research Reactor (TRR-1/M1)

International Nuclear Information System (INIS)

Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

2006-01-01

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

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)
Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

Energy Technology Data Exchange (ETDEWEB)

Montes, J

1972-07-01

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

International Nuclear Information System (INIS)

Perez, P.B.

1993-01-01

The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities
Dr. Anna G. Jonasdottir: Acceptance Speech for Honorary Doctorate from Faculty of Political Science, University of Iceland. Given 18th of June, 2015

OpenAIRE

Anna G. Jónasdóttir

2016-01-01

On June 18th the Faculty of Political Science, University of Iceland, celebrated 100 years of women’s‘ voting rights in Iceland with a special conference, Power and democracy 100 years later. In association with the conference Dr. Anna Guðrún Jónasdóttir, Professor emerita at the University of Örebro, Sweden, was awarded an honorary doctorate at the Faculty of Political Science. Anna Guðrún was the first Icelandic woman to complete a doctorate in political science, in 1991, and also the first...
The "Matchbox School" (1927-1932): Anna Freud and the Idea of a "Psychoanalytically Informed Education"

Science.gov (United States)

Midgley, Nick

2008-01-01

Of all the applications of psychoanalysis to various fields, perhaps none has been as important--or as fraught--as the application of psychoanalytic insights to education. This paper re-constructs some of the early debates around psychoanalysis and pedagogy that Anna Freud engaged with during the 1920s in Vienna, when the whole question of what…
EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

Energy Technology Data Exchange (ETDEWEB)

Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-07-01

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.
Annual report on JEN-1 reactor

International Nuclear Information System (INIS)

Montes, J.

1972-01-01

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

Directory of Open Access Journals (Sweden)

RINA MARTIARA

2013-11-01

Full Text Available Symbolic Analysis of Kyrie Eleison of Anna Wahyu Prasetyowati’s Work. This paper describes and interpretsthe Kyrie Eleison as a performing arts. Kyrie Eleison text is read through choreography elements that shape thethemes, motion, musical accompaniment, fashion makeup, fl oor pattern, form and stage layout. This paper useshermeneutic approach. Based on the research it can be concluded that this work has a deep meaning. The story ofJesus Christ journey in teaching the word with great passion until his death on the Cross is fi xated religious valuesthat serve as an example for mankind. This work also contains the values of humanity that every human being mustbe having a miserable, suffering, and death. Humans must be steadfast and strong in the face of every temptationthat comes when he/she want to achieve the highest inner awareness and understanding of divine truth.
Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

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

Costos y rendimientos de producción de tres néctares de manzana (pyrusmalusl variedades Anna, Pensilvania y Winter

Directory of Open Access Journals (Sweden)

Darío Alberto Pinto Medina

2016-12-01

Full Text Available La agroindustria en Colombia se muestra como una alternativa para dar aprovechamiento eficiente a materias primas de origen agropecuario, especialmente aquellas que por parámetros de calidad, no salen al mercado y generan cuantiosas pérdidas a agricultores y ganaderos; esta investigación se basó principalmente en el aprovechamiento de la manzana (Pyrus malus L, se elaboró néctar con tres variedades producidas en el departamento de Boyacá, Colombia: Pensilvania, Anna y Winter y se evaluaron los rendimientos en extracción de pulpa, con pérdidas del 18,4% para Pensilvania 21,4%, para Anna y 30%en Winter. Se realizaron pruebas sensoriales para determinar el grado de aceptabilidad de los tres productos, observándose diferencias significativas en textura, y similitud en parámetros como olor, color y sabor; para dicho análisis se compararon las tres variedades con un néctar comercial. Se concluye que el costo de producción unitario para 900 ml de néctar de la variedad Pensilvania es de $4.999, Anna $5.100 y Winter $5.296 (pesos colombianos, esto se debe a mejores rendimientos en extracción de pulpa para la variedad Pensilvania, mostrando que los productos son competitivos en precio y calidad.
Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

2003-01-01

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

Index Scriptorium Estoniae

2007-01-01

Vastavad Narva Kutseõppekeskuse kutseõpetajad Irina Burkina, Tatjana Dolgovskaja ja Murat Ibragimov ning autovaldkonna juht Kaarel Patlep, projektijuht Ljubov Kaalik ja PR-spetsialist Anna Vorobjova
Produção de conhecimento na enfermagem em oncologia: contribuição da escola de enfermagem Anna Nery Producción de conocimiento en la enfermería en oncología: contribución de la escola de enfermagem anna nery (escuela de enfermería Anna Nery Production of knowledge in nursing in oncology: contribution of Anna Nery nursing school

Directory of Open Access Journals (Sweden)

Marléa Chagas Moreira

2010-09-01

Full Text Available O estudo teve como objetivos mapear as teses e dissertações produzidas no Programa de Pós-Graduação da Escola de Enfermagem Anna Nery com enfoque na enfermagem em oncologia e analisar as repercussões diante dos aspectos epistemológicos destacados. Estudo exploratório, descritivo, retrospectivo e bibliográfico de 59 estudos (45 dissertações e 14 teses produzidos de 1980 a 2009. O referencial teórico-metodológico foi a Metodologia de Categorização Epistemológica para a Pesquisa na Enfermagem. Os resultados indicam que a produção analisada é mais setorizada na área da oncologia clínica, com ênfase nos aspectos Gestão Saúde/Enfermagem, Saúde da Mulher e Fundamentos do Cuidado de Enfermagem. Houve predomínio de estudos na perspectiva sociológica com abordagem qualitativa. As repercussões demonstradas nos aspectos epistemológicos destacados possibilitam afirmar que o conhecimento produzido é consistente com a complexidade e as tentativas de explicação sobre a arte de cuidar dos clientes com câncer e com esforços de definir/ampliar critérios e padrões assistenciais. Enfermagem Oncológica. Pesquisa em Enfermagem. Educação de Pós-graduação em EnfermagemEl estudio tuvo como objetivo mapear las tesis y disertaciones del Programa de Posgraduado de la Escola de Enfermagem Anna Nery (Escuela de Enfermería Anna Nery con énfasis en la enfermería en oncología y analizar las repercusiones frente los aspectos epistemológicos señalados. Exploratorio, descriptivo, retrospectivo y documental de 59 estudios (45 disertaciones y 14 tesis producidos desde 1980 hasta 2009. El referencial teórico-metodológico fue la Metodología Epistemológica para la Investigación en Enfermería. Los resultados indican que la producción es la más sectorizada de la oncología clínica, con énfasis en Gestión de la Salud / Enfermería, Salud de la Mujer y Fundamentos de la Atención de Enfermería. Se detectó un predominio de los
Homogeneous SLOWPOKE reactors for Mo-99/Tc-99m production in North America

Energy Technology Data Exchange (ETDEWEB)

Hilborn, J.W., E-mail: hilbovanw@sympatico.ca [Deep River, Ontario (Canada); Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)

2014-07-01

The 15 month shutdown of NRU in 2009 - 2010 caused an overall isotope shortage of approximately 30%; and in North America, the annual Tc-99m demand decreased from an estimated 20 million unit doses to about 15 million unit doses. Mo-99/Tc-99m is produced from HEU targets, irradiated in NRU for 11 days, and after chemical removal of uranium it is shipped to Nordion in Kanata, Ontario. Nordion further purifies the material and sends it to Lantheus Medical Imaging in the USA for manufacture of Mo-99 generators, which are then distributed to hundreds of hospital radiopharmacies throughout North America. One other American company, Covidien, manufactures and distributes Mo-99 generators like Lantheus, but they import bulk Mo-99 from Europe or South Africa. At the hospitals, Tc-99m is chemically extracted daily from the Mo-99 generators and loaded into syringes for immediate clinical use. Fortuitously, the 66 hour half-life of Mo-99 allows the replenishment of Tc-99m in the generator over a growth period of about 20 hours; and a generator can be 'milked' daily for up to two weeks. A more efficient model is the direct production and distribution of Tc-99m unit doses to regional hospitals from 10 'industrial' radiopharmacies located at existing licensed reactor sites in North America. A 20 kW homogeneous SLOWPOKE reactor at each site would deliver 15 litres of irradiated uranyl sulphate fuel solution daily to industrial-scale hot cells for extraction of Mo-99, which would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and the Low Enriched Uranium (LEU) would be recycled. Each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily, for courier delivery to all of the Nuclear Medicine hospitals within a 3 hour average range by road transport. Typically, the delivered doses would be in the range 10 to 30 mCi. Assuming an average unit dose of 25 mCi at the hospital and 5 x 52
Moving ring reactor 'Karin-1'

International Nuclear Information System (INIS)

1983-12-01

The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)
The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)
Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented
Patronage in the dispute over child analysis between Melanie Klein and Anna Freud--1927-1932.

Science.gov (United States)

Aguayo, J

2000-08-01

The author investigates the role of patrons and advocates for Melanie Klein's clinical ideas at the British Psycho-Analytical Society against the backdrop of her theoretical and technical differences with Anna Freud from 1927 to 1932. He also outlines the development of Klein and Anna Freud's theories and techniques within the nascent discipline of child psychoanalysis. The London and Viennese patrons/advocates contributed to polarising what initially were clinical differences about how to analyse pre-latency and latency-age children and which technical processes might best facilitate successful treatment. While the author speculates that a diversity of motivations and agendas may have driven the London group's support for Klein--personal and politicised enthusiasm (Jones), genuine conviction (Riviere) and attempts at theoretical rapprochement between the London and Vienna schools (Glover)--he also argues that Freud's diagnosis with cancer in 1923 and suspicion of patricidal son-successors necessitated the choice of a female successor with unquestioning loyalty to his doctrines. From 1932, when Klein's clinical authority was established, her first group of English supporters began to splinter, as she went on to become a training analyst, mentor and patron in her own right to a succeeding generation of adherents who defended her views during the Controversial Discussions.
Occupational analysis for the Angra-1 reactor

International Nuclear Information System (INIS)

Moraes, A.

1991-01-01

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

Science.gov (United States)

Magno, JoJo

2009-01-01

In attempting to climb past the racist and sexist barriers which existed in nineteenth-century America, women could look to writers such as Harriet Beecher Stowe and Anna Julia Cooper. Their works not only reflect the conditions of women and African-American women in particular, but also call for access to educational opportunities for these women…
Dr. Anna G. Jonasdottir: Acceptance Speech for Honorary Doctorate from Faculty of Political Science, University of Iceland. Given 18th of June, 2015

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Anna G. Jónasdóttir

2016-06-01

Full Text Available On June 18th the Faculty of Political Science, University of Iceland, celebrated 100 years of women’s‘ voting rights in Iceland with a special conference, Power and democracy 100 years later. In association with the conference Dr. Anna Guðrún Jónasdóttir, Professor emerita at the University of Örebro, Sweden, was awarded an honorary doctorate at the Faculty of Political Science. Anna Guðrún was the first Icelandic woman to complete a doctorate in political science, in 1991, and also the first to embark on an advanced academic career in political science and gender studies. It is therefore highly appropriate that Anna Guðrún should be awarded the first honorary doctorate at the Faculty of Political Science, where these disciplines are located. Her research covers a broad spectrum, including political science, sociology, economic history, psychology and gender studies. She was among the first to deal in a theoretical manner with gender, power and politics, which was considered rather provocative at the start of her academic career in the early 1970s. She is a pioneer in intertwining political research and gender studies and her most important research is in the field of power and personal gender relations. Anna Guðrún moved to Sweden at an early age but has kept in touch with the Icelandic research community. Below we publish her acceptance speech on the occasion when the honorary doctorate was awarded. It reflects clearly how her ideas have developed and her intimate sense for how personal and political factors bring politics and gender studies closer at the same time as she deepens and broadens both of their subjects.
["Signs for the whole rest of the life". The house as symbol and autobiographical substrate in letters, dreams and literary texts by by Anna Freud].

Science.gov (United States)

Spreitzer, Brigitte

2015-01-01

The house as object and symbol preoccupied Anna Freud from childhood onwards. This article traces these ideas in letters, autobiographical documents and literary texts. It focuses on the phantasmal construction of a dream house; on the organisation, loss and later efforts to recover Hochrotherd; on casual birthday poems dedicated to Anna, praising the country house, as well as on the purchase of a weekend cottage in Walberswick reflected in dreams related to the process of inner detachment from the dead father. A fluent interchange between reality and imagination comes into focus, high- lighting the dimension of inwardness rather than that of biographical reality.
Reactor operations at SAFARI-1

International Nuclear Information System (INIS)

Vlok, J.W.H.

2003-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Takac, S M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1972-07-01

The method is based on perturbation of the reactor cell from a few up to few tens of percent. Measurements were performed for square lattice calls of zero power reactors ANNA, NORA and RB, with metal uranium and uranium oxide fuel elements, water, heavy water and graphite moderators. Character and functional dependence of perturbations were obtained from the experimental results. Zero perturbation was determined by extrapolation thus obtaining the real physical neutron flux distribution in the reactor cell. Simple diffusion theory for partial plate cell perturbation was developed for verification of the perturbation method. The results of these calculation proved that introducing the perturbation sample in the fuel results in flattening the thermal neutron density dependent on the amplitude of the applied perturbation. Extrapolation applied for perturbed distributions was found to be justified.
Review Essay: John Pratt and Anna Eriksson (2013 'Contrasts in Punishment: An Explanation of Anglophone Excess and Nordic Exceptionalism'

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David Brown

2013-11-01

Full Text Available This review essay combines the comments made by David Brown, Russell Hogg and Mark Finanne at the Crime, Justice and Social Democracy: 2nd International Conference July 2013. It is followed by a rejoinder by the two authors John Pratt and Anna Eriksson.
Die Zitruskultur am Hofe Ferdinands I. und Anna Jagiellos. Import und Anbau von Südfrüchten in Prag 1526–1564

Czech Academy of Sciences Publication Activity Database

Dobalová, Sylva; Hausenblasová, J.

-, č. 15 (2015), s. 9-36 ISSN 1213-5372 Institutional support: RVO:68378033 Keywords : Ferdinand I * Anna of Jagiello * orange fruit * renaissance gardens * trade * residence Subject RIV: AL - Art, Architecture, Cultural Heritage
Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Karel, Matejka; Lubomir, Sklenka

2005-01-01

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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Anna Sriastuti

2017-04-01

Full Text Available By not giving voice and value to women‘s opinions, responses and writings, men have therefore suppressed the female, define what it means to be feminine, and therefore devoiced, devalued, and trivialized what it means to be a woman. As femininity is mostly related to women for women are labeled as extensions of men, mirrors of men, devices for showing men off, and also devices for helping men get what they want, women‘s position is inferior to men. This study discusses the female‘s masculinities through the portrayal and life experiences of the main female characteristics, Anna Leonowens, as her protests against the Siamese patriarchal systems that abundant her life as a career woman and also the oppressed lives of Siamese women in Elizabeth Hand‘s novel Anna and the King. Feminism and Deconstructions approaches will be applied to analyze Leonowens‘ actions and reactions regarding to what Siamese patriarchal systems claim to be parts of masculine traits. Leonowens‘ nationality, cultural and educational backgrounds as well as positions play significant roles in the novel. By opposing Leonowens‘ ways of thinking to the King who represents the patriarchal system of Siam, Hand gives a new identity to women in struggling for gender equality in Siam.

Analisando o conteúdo da seção "Página do Estudante" dos Annaes de Enfermagem Analisando el contenido de la sección "Pagina del Estudiante" de los Annaes de Enfermagem Analysing the content of the section Students' page of the Annaes de Enfermagem

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Marlene Nunes Morais Pereira

2006-01-01

Full Text Available A Revista Annaes Enfermagem foi um veiculo para retratar os problemas da saúde e evidenciar o exercício da profissão. A seção página do estudante foi criada para dar incentivo ao aluno, para expor suas idéias e focar pontos importantes na formação do enfermeiro da época. Este trabalho teve por objetivo analisar e descrever o conteúdo das publicações no período de 1932 a 1941. Trata-se de estudo de abordagem qualitativa pelo método histórico. Os principais elementos descritos na seção analisada estavam relacionados com os problemas de saúde, como escarlatina, tétano eclâmpsia. Conclui-se que o conteúdo do material analisado reflete que neste período existia uma preocupação com a preparação das alunas para suprir a necessidade vigente que o país atravessava.El periodico Annaes de Enfermagem ha sido el veiculo para discutir los problemas de salud y para evidenciar el professional rol. La sección "Pagina de estudiante" há sido creada para incentivar los estudiantes a exponer sus ideas y para focalisar en los puntos relevantes de formación del enfermero el aquelle tienpo . Esto estúdio objectivó describir y analisar el contenido de lo material publicado em el periodo de 1932 hasta 1941. Esto és um estudio con abordaje cualitativo usandoce el metodo historìco. Los principales elementos descriptos en la sección analisada estan relacionados a los problemas de salud colectiva como la escarlatina, tétano, eclampsia y tanbien a la ética. Se ha concluydo que el contenido del material analisado reflete que en el período analisado habia una preocupación com la preparación del studiante para atender a las necesidades de salud que el pais presentava.The journal Annaes de Enfermagem was a publishing vehicle to discuss health problems and to evidence the professional role. The section "Students Page" Was created for the motivation of students to expose their ideas and to focus in relevant points of the nurse education by
How much plutonium does North Korea really have?

International Nuclear Information System (INIS)

Dreicer, J.S.

1997-01-01

In a previous study, as part of the Global Nuclear Material Control Model effort, the author estimated the maximum quantity of plutonium that could be produced in thermal research reactors in the potential nuclear weapon states (including North Korea), based on their declared power level. D. Albright has estimated the amount of plutonium the North Koreans may have produced since 1986 in the 5-megawatt-electric power reactor at Yongbon. Albright provided an upper-bound estimate of 53 kilograms of weapon-grade plutonium produced cumulatively if the gas-graphite (magnox) reactor had achieved a load factor of 0.80. This cumulative estimate of 53 kilograms ignores the potential plutonium production in the 8-megawatt-thermal research reactor, IRT-DPRK. To better quantify the cumulative North Korean production, the author conducted a study to estimate the amount of plutonium that could have been produced in the IRT-DPRK research reactor operating at the declared power level during the entire period it has operated, including a period it was not safeguarded. The author estimates that, at most, an additional 6 to 33 kilograms of plutonium could have been produced cumulatively in the research reactor operating at the declared power level during the entire period it has operated, including a 12-year period it was not safeguarded, resulting in a total of 13 to 47 kilograms of plutonium possibly produced in both the research and power reactors
Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Czech Academy of Sciences Publication Activity Database

Wögerbauer, Michael

2015-01-01

Roč. 39, č. 2 (2015), s. 240-243. ISBN 978-3-8353-16-97-3. ISSN 0722-740X Institutional support: RVO:68378068 Keywords : enlightenment * prose * Sager, Maria Anna * female writer * German literature * Czech literature Subject RIV: AJ - Letters, Mass-media, Audiovision
Physiatrie and German maternal feminism: Dr. Anna Fischer-Dückelmann critiques academic medicine.

Science.gov (United States)

Meyer, Paulette

2006-01-01

Alternative medicine and reform strategies made Anna Fischer-Dückelmann a most controversial, notorious, and widely read women doctor before World War I. She published a dozen titles in 13 languages asserting that national well-being depended on maternal prowess. To her critics, Fischer-Dückelmann's commitment to medical self-help and practices of Physiatrie amounted to medical quackery. Her career has been largely unexamined, yet her feminist critiques and social concerns are not far removed from modern social medicine. For this pioneering doctor, treating physical and emotional ills and promoting the health of families were first steps toward healing the divisions of a world at war.
Outline of construction planning on No. 2 Reactor of the Shika Nuclear Power Plant

International Nuclear Information System (INIS)

Nakagawa, Tetsuro; Kadoki, Shuichi; Kubo, Tetsuji

1999-01-01

The Hokuriku Electric Co., Ltd. carries out the expansion of the Shika Nuclear Power Plant No.2 (ABWR) to start its in March 2006. It is situated in north neighboring side of No. 1 reactor under operation at present, and its main buildings are planned to position a reactor building at mountain side and a turbine building at sea side as well as those in the No. 1 reactor. And, cooling water for steam condenser was taken in from an intake opening built at north side of the lifting space situated at the front of the power plant, and discharged into seawater from a flashing opening positioned about 600 m offing. Here were described on outline of main civil engineering such as base excavation engineering, concrete caisson production, oceanic establishment engineering, and facility for steam condenser, and characteristics of the engineering. (G.K.)
Astronaut Anna Fisher Suits Up for NBS Training

Science.gov (United States)

1980-01-01

The Hubble Space Telescope (HST) is a cooperative program of the European Space Agency (ESA) and the National Aeronautical and Space Administration (NASA) to operate a long-lived space-based observatory. It was the flagship mission of NASA's Great Observatories program. The HST program began as an astronomical dream in the 1940s. During the 1970s and 1980s, the HST was finally designed and built becoming operational in the 1990s. The HST was deployed into a low-Earth orbit on April 25, 1990 from the cargo bay of the Space Shuttle Discovery (STS-31). The design of the HST took into consideration its length of service and the necessity of repairs and equipment replacement by making the body modular. In doing so, subsequent shuttle missions could recover the HST, replace faulty or obsolete parts and be re-released. Marshall Space Flight Center's (MSFC's) Neutral Buoyancy Simulator (NBS) served as the test center for shuttle astronauts training for Hubble related missions. Shown is astronaut Anna Fisher suiting up for training on a mockup of a modular section of the HST for an axial scientific instrument change out.
Astronaut Anna Fisher Suiting Up For NBS Training

Science.gov (United States)

1980-01-01

The Hubble Space Telescope (HST) is a cooperative program of the European Space Agency (ESA) and the National Aeronautical and Space Administration (NASA) to operate a long-lived space-based observatory. It was the flagship mission of NASA's Great Observatories program. The HST program began as an astronomical dream in the 1940s. During the 1970s and 1980s, the HST was finally designed and built becoming operational in the 1990s. The HST was deployed into a low-Earth orbit on April 25, 1990 from the cargo bay of the Space Shuttle Discovery (STS-31). The design of the HST took into consideration its length of service and the necessity of repairs and equipment replacement by making the body modular. In doing so, subsequent shuttle missions could recover the HST, replace faulty or obsolete parts and be re-released. Marshall Space Flight Center's (MSFC's) Neutral Buoyancy Simulator (NBS) served as the test center for shuttle astronauts training for Hubble related missions. Shown is astronaut Anna Fisher suiting up for training on a mockup of a modular section of the HST for an axial scientific instrument change out.
Astronaut Anna Fisher Suited Up For NBS Training

Science.gov (United States)

1980-01-01

The Hubble Space Telescope (HST) is a cooperative program of the European Space Agency (ESA) and the National Aeronautical and Space Administration (NASA) to operate a long-lived space-based observatory. It was the flagship mission of NASA's Great Observatories program. The HST program began as an astronomical dream in the 1940s. During the 1970s and 1980s, the HST was finally designed and built becoming operational in the 1990s. The HST was deployed into a low-Earth orbit on April 25, 1990 from the cargo bay of the Space Shuttle Discovery (STS-31). The design of the HST took into consideration its length of service and the necessity of repairs and equipment replacement by making the body modular. In doing so, subsequent shuttle missions could recover the HST, replace faulty or obsolete parts and be re-released. Marshall Space Flight Center's (MSFC's) Neutral Buoyancy Simulator (NBS) served as the test center for shuttle astronauts training for Hubble related missions. Shown is astronaut Anna Fisher suited up for training on a mockup of a modular section of the HST for an axial scientific instrument change out.
Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

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

1994-01-01

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

International Nuclear Information System (INIS)

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

2002-01-01

The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)
Anna-Caterina Walk: Das Andere im Tatort. Migration und Integration im Fernsehkrimi. Marburg: Tectum Wissenschaftsverlag 2011.

Directory of Open Access Journals (Sweden)

Patricia Piberger

2012-08-01

Full Text Available Aus medien- und kulturwissenschaftlicher Perspektive widmet sich Anna-Caterina Walk in ihrer knappen Publikation der TV-Krimireihe ‚Tatort‘. Anhand von drei konkreten Folgen analysiert sie, wie das Andere in dieser Serie medial repräsentiert und konstruiert wird. Ihr Ziel ist es, die konkreten Darstellungen und deren Bedeutungen in ihrer Selbstverständlichkeit zu hinterfragen. Dabei sucht sie vor allem nach nicht-stereotypen oder destabilisierenden Bildern und fordert zugleich einen kritischeren Umgang mit Differenzkonstruktionen der Identität innerhalb der einzelnen ‚Tatort‘-Folgen. Zumal der Fokus der Autorin auf kulturellen Andersartigkeiten liegt, kommen Betrachtungen des_der geschlechtlich Anderen eine marginale Position zu und bleiben prinzipiell eher oberflächlich.In her concise publication, Anna-Caterina Walk dedicates herself to the TV crime series Tatort using a media studies and cultural studies perspective. Based on three specific episodes, she analyzes how the ‘other’ is medially represented and constructed in the series. She aims at questioning the self-evidence of specific representations and their meanings. In doing so, she searches, above all, for non-stereotypical or destabilizing images and, at the same time, demands a more critical approach to constructions of difference regarding identity within the individual Tatort episodes. Considerations of the gendered ‘other/s’ are treated only marginally and generally remain superficial, particularly since the author focuses on cultural otherness.
End effects in the criticality analysis of burnup credit casks

International Nuclear Information System (INIS)

Brady, M.C.; Parks, C.V.

1990-01-01

A study to evaluate the effect of axially dependent burnup on k eff has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 x 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % 235 U and an average burnup of 31.5 GWd/MTU
Training and research on the nuclear reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

1998-01-01

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

International Nuclear Information System (INIS)

Trevena, I.

1994-01-01

Reactors have played a key pole in the production of radioactive isotopes for medical applications for the past 50 years. This paper reviews current and future capabilities for the production and supply of radioactive isotopes used in nuclear medicine. It focuses primarily on the supply of fission product molybdenum-99, which is used to produce technetium-99m, the radioisotope most widely employed in nuclear medicine procedures. The significant infrastructure required for the production and supply of molybdenum-99 is detailed, and the capabilities of the major commercial suppliers in North America and Europe are discussed. Plans for increasing production capabilities in the future are also reviewed. (author)
Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

Otra Otra, F; Leira Rey, G

1971-07-01

In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.
IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs
[Questions and worries. On the correspondence of Grete Bibring and Anna Freud 1949-1975].

Science.gov (United States)

Bakman, Nina

2015-01-01

Grete Bibring (1899 - 1977) was a representative of the second generation of analysts. Having emigrated from Vienna to London in 1938, she left for Boston in 1942 where she made a remarkable career. 1946 she became head of the department of psychiatry at the Beth-Israel hospital in Harvard and from 1961 the first woman professor of medicine there. She maintained a connection with European psychoanalysis in the person of Anna Freud with whom she corresponded regularly. Their letters contain an interesting exchange of ideas about psychoanalytic institutions (e.g. the American Psychoanalytical Association) and papers (e.g. on pregnancy). It is also the testimony of an exceptional friendship.
TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

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

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.
TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

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

1994-01-01

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

Modernization and Refurbishment of the RECH-1 Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

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

International Nuclear Information System (INIS)

Rataj, J.

2010-01-01

The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo
Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

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

Directory of Open Access Journals (Sweden)

Mela Cristanty

2016-12-01

Full Text Available This study discusses the interpersonal communication between the nurses with elderly who were living in a nursing home. Establishment of a nursing home is one of the government's efforts to address the issue of social welfare for the elderly. Like the other nursing homes Nursing Home “Santa Anna” also have nurses on duty to take care the elderly living at home. In order to perform their duties the nurses need to perform an effective interpersonal communication with the elderly. Interpersonal communication activities have influence in building a relationship with the elderly in order to provide guidance, supervision and encouragement to the elderly. The theory used in this research is theory of communication, interpersonal communication, social penetration theory, concept of nurses, the elderly and nursing homes. This study used descriptive qualitative method and authors conducted in-depth interviews with four informants, namely two nurses and two elderly people, also made observation and doing literature study. The results of this study reveals that the closeness of interpersonal relations between nurses and elderly in Nursing Home “Santa Anna” can be seen through five general quality that are openness, positive behavior, supportive behavior, empathy, and equality. These five factors mentioned above is run entirely by nurses in the Nursing Home “Santa Anna” in terms of communicating and forming relationships with the elderly who live in institutions. Penelitian ini membahas mengenai komunikasi antarpribadi yang dilakukan antara perawat dengan lansia yang tinggal di panti jompo. Pendirian panti jompo merupakan salah satu usaha pemerintah dalam menangani masalah kesejahteraan sosial untuk para lansia. Seperti panti-panti jompo lainnya Panti Lansia Santa Anna juga memiliki perawat-perawat yang bertugas untuk merawat, mengurus, dan mengasuh para lansia yang tinggal di panti. Untuk dapat menjalankan tugas serta perannya perawat perlu
Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

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

International Nuclear Information System (INIS)

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

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs
Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

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

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels
Operation and maintenance of 1MW PUSPATI TRIGA reactor

International Nuclear Information System (INIS)

Adnan Bokhari; Mohammad Suhaimi Kassim

2006-01-01

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

International Nuclear Information System (INIS)

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

1997-11-01

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-11-01

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

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

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

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

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

International Nuclear Information System (INIS)

Mosteller, R.D.

1996-01-01

The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE's geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs)
Validation of NESTLE against static reactor benchmark problems

International Nuclear Information System (INIS)

Mosteller, R.D.

1996-01-01

The NESTLE advanced nodal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE's geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs), and Canada deuterium uranium (CANDU) heavy-water reactors (HWRs)
Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

International Nuclear Information System (INIS)

Perry, E.F.

1995-01-01

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

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1981-01-01

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

International Nuclear Information System (INIS)

Perez, P.B.

1997-01-01

The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996
Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

International Nuclear Information System (INIS)

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

2002-01-01

Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)
RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

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

1985-01-01

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

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

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

Science.gov (United States)

2007-04-18

Special Report (1). The Economist, May 3, 2003 (U.S. Edition). 13 See, for example: Natural Wonders Prove Kim Jong-Il’s Divinity: North Korean Media ...clams and matsutake mushrooms are particularly prized in the Japanese market. Japan sends items such as vehicles, electrical machinery, boilers/reactors...client of BDA was widely reported to be conducting “numerous illegal activities, including distributing counterfeit currency and smuggling counterfeit
Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was
Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

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

International Nuclear Information System (INIS)

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

2002-01-01

The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)
Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

International Nuclear Information System (INIS)

1988-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)
Department of Reactor Technology annual progress report 1 January -31 December 1977

International Nuclear Information System (INIS)

1978-04-01

The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)
Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

International Nuclear Information System (INIS)

1989-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)
Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

International Nuclear Information System (INIS)

Ratanatongchai, Wichian

2009-07-01

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

International Nuclear Information System (INIS)

Kropik, M.; Jurickova, M.

2006-01-01

The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)
Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

International Nuclear Information System (INIS)

1991-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)
Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

International Nuclear Information System (INIS)

1992-08-01

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

Directory of Open Access Journals (Sweden)

Nadja Louwerse

2010-03-01

Full Text Available Dutch nautical terms in Russian. Anna Croiset van der Kop versus Reinder van der Meulen This article explores the controversy in the beginning of the 20th century between de Dutch slavists Anna Croiset van der Kop and Reindert van der Meulen, author of the book De Hollandsche Zee- en Scheepstermen in het Russisch (Dutch Nautical Terms in Russian. After publication, she reacted with an undeniably negative and unusually elaborate review (72 pages!, written not in Dutch, but in Russian and published in one of the most leading, scientific series in Russia. Van der Meulen has never directly reacted to this review: in 1959 he published a supplement to his earlier work from 1909: Nederlandse woorden in het Russisch (Dutch words in Russian, not mentioning Croiset van der Kops name, nor her review. Obviously he felt hurt by her merciless criticism. The article focuses on the motives underlying her strong reaction. It will be argued that her severe criticism was not really inspired by Van der Meulen’s subject, but by his failing research methods. Furthermore, it will be pointed out that this review seemed to offer her an opportunity to measure up to her Dutch colleagues and to present herself internationally as a slavist. On the basis of the now available data it is hard to say, which motives really brought Croiset van der Kop to write her review. Investigation of her archives, preserved in St- Petersburg, and until now hardly studied, may perhaps reveal her real motives.
Astronaut Anna Fisher in NBS Training For Hubble Space Telescope

Science.gov (United States)

1980-01-01

The Hubble Space Telescope (HST) is a cooperative program of the European Space Agency (ESA) and the National Aeronautical and Space Administration (NASA) to operate a long-lived space-based observatory. It was the flagship mission of NASA's Great Observatories program. The HST program began as an astronomical dream in the 1940s. During the 1970s and 1980s, the HST was finally designed and built becoming operational in the 1990s. The HST was deployed into a low-Earth orbit on April 25, 1990 from the cargo bay of the Space Shuttle Discovery (STS-31). The design of the HST took into consideration its length of service and the necessity of repairs and equipment replacement by making the body modular. In doing so, subsequent shuttle missions could recover the HST, replace faulty or obsolete parts and be re-released. Marshall Space Flight Center's (MSFC's) Neutral Buoyancy Simulator (NBS) served as the test center for shuttle astronauts training for Hubble related missions. Shown is astronaut Anna Fisher training on a mock-up of a modular section of the HST for an axial scientific instrument change out.
Rec.: Anna Niepytalska-Osiecka, Socjolekt polskich alpinistów. Analiza leksykalno-semantyczna słownictwa, Wydawnictwo LIBRON – Filip Lohner, Kraków 2014, ss. 286

Directory of Open Access Journals (Sweden)

Beata Jarosz

2017-12-01

Full Text Available Review: Anna Niepytalska-Osiecka, Socjolekt polskich alpinistów. Analiza leksykalnosemantyczna słownictwa, Wydawnictwo LIBRON – Filip Lohner, Kraków 2014, 286 pp. The article is a review of Anna Niepytalska-Osiecka’s book Socjolekt polskich alpinistów. Analiza leksykalno-semantyczna słownictwa (The sociolect of Polish mountaineers. A lexical and semantic analysis of the vocabulary. First, the text summarizes the subject matter of this monograph, devoted to a hitherto undescribed variety of the Polish language. It then goes on to characterize the composition of the book and its methodology. Finally, some problematic points are indicated – concerning, among other things, the classification of the language variety analyzed or the development of its vocabulary – and alternative lexical and semantic solutions are suggested. Rec.: Anna Niepytalska-Osiecka, Socjolekt polskich alpinistów. Analiza leksykalno-semantyczna słownictwa, Wydawnictwo LIBRON – Filip Lohner, Kraków 2014, ss. 286 Artykuł stanowi recenzję książki Anny Niepytalskiej-Osieckiej pt. Socjolekt polskich alpinistów. Analiza leksykalno-semantyczna słownictwa. W tekście przybliżono zawartą w monografii problematykę poświęconą nieopisanemu dotychczas wariantowi polszczyzny oraz scharakteryzowano zastosowane rozwiązania kompozycyjne i analityczne. Wskazano również punkty dyskusyjne dotyczące m.in. klasyfikacji opisywanego wariantu polszczyzny czy procesu kształtowania się jego zasobu słownego, a także zasugerowano alternatywne rozstrzygnięcia leksykalno-semantyczne.
Two Austro-Hungarian Women Writers, Anna Tutsek and Terka Lux, Creating New Urban Identities in Early Twentieth-Century Budapest

OpenAIRE

Judit Kádár

2016-01-01

In this paper, I examine some literary texts of two turn-of-the century Hungarian women writers, Anna Tutsek and Terka Lux who left behind their childhood environment in remote regions of the Austro-Hungarian Monarchy in order to move to Budapest, the capital of the eastern part of the Empire. Assuming that individuals hold multiple identities that are flexible and inevitably affected by environmental and social changes, my main focus is on the transformation of their ethnic, regional, occupa...
Thermohydraulic analysis for power increase of IEAR-1 reactor

International Nuclear Information System (INIS)

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

1996-01-01

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

Directory of Open Access Journals (Sweden)

Ekaterina Gurina

2016-09-01

Full Text Available The article deals with the stylistic peculiarities of the translations made by R. Pevear and L. Volokhonsky, L. and A. Maud, J. Carmichael of the novel Anna Karenina by L. Tolstoy on the basis of pragmastylistics and comparative analysis. It tries to analyze the text of the novel using the lingo-stylistic characteristics in accordance with the national bias in the way of thinking and individual creative preferences of every translator taking an attempt to introduce a foreign picture of the world to his countrymen. It underlines the impact of Tolstoy’s complicated attitude towards the customs and traditions of the Russian Orthodox church and the specific relationship of the author of the novel with God and its manifestation in the description of the heroes’ characters. In stresses how vital it may turn out to preserve the author’s ideostyle - lexis and syntax (the word order, the choice of them and the length of the sentences for the successful interpretation of the writer’s views and stance by the reader.
Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)
Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

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

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

International Nuclear Information System (INIS)

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

1981-12-01

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The
76 FR 54259 - Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station...

Science.gov (United States)

2011-08-31

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0185] Virginia Electric and Power Company, Docket Nos. 50.... NPF-4 and NPF-7, issued to Virginia Electric Power Company (the licensee), for operation of the North...) and (d) during declarations of severe weather conditions involving tropical storm or hurricane force...
Department of Reactor Technology: annual progress report 1 January - 31 December 1976

International Nuclear Information System (INIS)

1977-06-01

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

International Nuclear Information System (INIS)

Lee, Jung Hyu; An, Jin Soo

2011-01-01

On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor
Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)
Neutron density optimal control of A-1 reactor analoque model

International Nuclear Information System (INIS)

Grof, V.

1975-01-01

Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)
Department of Reactor Technology annual progress report 1 January - 31 December 1978

International Nuclear Information System (INIS)

1979-04-01

The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)
Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

International Nuclear Information System (INIS)

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)
Reactor engineering department annual report. April 1, 1993-March 31, 1994

International Nuclear Information System (INIS)

1994-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)
Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)
Reactor engineering department annual report. April 1, 1994 - March 31, 1995

International Nuclear Information System (INIS)

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)
Modification of the IAN-R1 reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

Science.gov (United States)

Hartman, John J

2014-12-01

This article explores the period of Anna Freud's life after she was informed of the deaths of her aunts in Nazi concentration camps during World War II. Understanding of this period may be enhanced by consideration of the role of the Holocaust in her complicated mourning process. A series of her dreams is re-examined from the point of view of survivor guilt and the complicated mourning of her father in the context of the Holocaust. It is argued that unconscious reproaches against her father led to an identification with him that included his 'decision' to leave his sisters in Vienna. Survivor guilt in relation to her aunts' murders is seen as one of the complicating factors in the mourning process. In addition the article discusses the possible role of this period, particularly her work with child concentration camp survivors, in her post-war writing. The noted duality in her work between innovation and conservatism is explored in terms of an outcome of the mourning process of this period. It is argued that her views on mourning, trauma, attachment, and the widening scope of indications for psychoanalysis were influenced by the outcome of her mourning process. Finally, an irony is noted in the fact that her attitude about altruism never changed despite the role of the altruism of others in her rescue from the Nazis. Copyright © 2014 Institute of Psychoanalysis.
Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

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

1986-01-01

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

International Nuclear Information System (INIS)

Kreuzer, W.; Hecht, H.

1988-01-01

The Cs-137 contamination of the soil in South Germany, especially around Schwabmuenchen, after the reactor accident in Chernobyl at the end of April 1986 amounted up to 20000 Bq/2. At certain places, maximum loads of even 40000 Bq/2 were measured. In the other South Bavarian regions and the southern parts of East Bavaria Cs-137 loads of between 5000-10000 Bq/2 were recorded which gradually declined to the North and to the West and reached values of [de
Burst muscle performance predicts the speed, acceleration, and turning performance of Anna's hummingbirds.

Science.gov (United States)

Segre, Paolo S; Dakin, Roslyn; Zordan, Victor B; Dickinson, Michael H; Straw, Andrew D; Altshuler, Douglas L

2015-11-19

Despite recent advances in the study of animal flight, the biomechanical determinants of maneuverability are poorly understood. It is thought that maneuverability may be influenced by intrinsic body mass and wing morphology, and by physiological muscle capacity, but this hypothesis has not yet been evaluated because it requires tracking a large number of free flight maneuvers from known individuals. We used an automated tracking system to record flight sequences from 20 Anna's hummingbirds flying solo and in competition in a large chamber. We found that burst muscle capacity predicted most performance metrics. Hummingbirds with higher burst capacity flew with faster velocities, accelerations, and rotations, and they used more demanding complex turns. In contrast, body mass did not predict variation in maneuvering performance, and wing morphology predicted only the use of arcing turns and high centripetal accelerations. Collectively, our results indicate that burst muscle capacity is a key predictor of maneuverability.
New treatment options for severe Arvophillia

International Development Research Centre (IDRC) Digital Library (Canada)

Carrie-Anne Whyte

Adaptation practice and policy. Climate change hotspots in South and Central Asia and Africa. Arctic North Consulting, 2012. Our role: 1. methodological consultants. 2. systematic review of adaptation practice and policy literature. Our team: • James Ford. • Lea Berrang –Ford. • Anna Bunce. • Maya Irwin. • Courtney McKay ...
Advanced Test Reactor National Scientific User Facility Partnerships

International Nuclear Information System (INIS)

Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

2012-01-01

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin
RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Jumel, Stephanie; Van-Duysen, Jean Claude

2005-01-01

Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program
Reactor engineering department annual report. April 1, 1995 - March 31, 1996

International Nuclear Information System (INIS)

1996-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).
Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

Energy Technology Data Exchange (ETDEWEB)

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

1998-11-01

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

International Nuclear Information System (INIS)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)
Reactor engineering department annual report. April 1, 1995 - March 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

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

International Nuclear Information System (INIS)

Festinesi, A.

1986-01-01

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

International Nuclear Information System (INIS)

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

1987-01-01

In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)
76 FR 17715 - Virginia Electric and Power Company North Anna Power Station, Units 1 and 2; Exemption

Science.gov (United States)

2011-03-30

... Optimized ZIRLO TM as a fuel cladding material based on: (1) Similarities with ZIRLO TM , (2) demonstrated... cladding material at NAPS. This change to the plant configuration has no relation to security issues... determined that the granting of this exemption will not have a significant effect on the quality of the human...
Elementos do Projeto Político Profissional da Associação Nacional das Enfermeiras Diplomadas Brasileiras presentes nos Annaes de Enfermagem Elementos del Proyecto Politico Profesional de la Asociación Nacional de Enfermeras Graduadas Brasileras presentes en los Annaes de Enfermagem Elements of the Political Professional Project of the Brazilian Graduate Nurses National Association present in the Annaes de Enfermagem

Directory of Open Access Journals (Sweden)

Jane Liliane Gonçalves da Cruz

2006-01-01

Full Text Available A revista Annaes de Enfermagem foi criada para publicar idéias, conceitos, resultados da produção científica, reflexões e, principalmente, o de expor e discutir o Projeto Político Profissional da Associação Nacional de Enfermeiras Diplomadas Brasileiras (ANEDB. Este estudo teve como objetivo descrever e caracterizar estes elementos presentes na referida revista no período de 1932 a 1941. Trata-se de um estudo de abordagem qualitativa pelo método histórico. Os principais elementos do projeto político compreendem os atributos desejáveis no enfermeiro como altruísmo, abnegação, patriotismo, senso humanitário, progresso profissional ligado à educação, arte, ideal, ética e componentes da religião cristã. Todos estes elementos fizeram parte de um projeto político que visava a integração da enfermagem no contexto de saúde nacional da época. (118 palavras.El periódico Annaes de Enfermagem fue creado para publicar ideas, conceptos, resultados de la producción cientifica, reflexiones y principalmente o de presentar y discutir el proyecto politico profesional de la Asociación Nacional de Enfermeras Graduadas Brasileras. Este estudio tuvo como objectivo decribir y caracterizar estos elementos presentes en este periódico en lo periodo de 1932 hasta 1941. Tratase de um estudio de abordaje cualitativa por lo metodo historico. Lo principales elementos del proyecto politico compreenden los atributos desejables en lo enfermero como el altruismo, abnegación, patriotismo, senso humaño, progreso profesional junto com la educación, arte, ideal, etica y componentes de la religión cristiana. Todos los elementos hicierón parte del proyecto politico que objectivó la integración de la enfermería en el contexto de salud en ámbito nacional a la epoca.The journal Annaes de Enfermagem was created to publish ideas, concepts, results of scientific production, reflections and, meanly to expose the Political Professional Project of the
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
Reliabitity study of the accumulator system for Angra-1 reactor

International Nuclear Information System (INIS)

Santos Maciel, C.C.R.

1980-01-01

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Directory of Open Access Journals (Sweden)

Igor Ximenes Graciano

2012-11-01

Full Text Available As narrativas de Sérgio Sant’Anna sugerem, pela exposição recorrente dos bastidores da criação literária (aqui entendida como gesto literário, que o discurso ficcional extrapola o espaço da literatura, repercutindo as relações dos indivíduos no “mundo real”. Em sua prosa é evidente a tentativa dos narradores de aproximarem o leitor para suas perspectivas, quando a narrativa assemelha-se a um jogo em que o prêmio é a cumplicidade deste a quem ela se dirige. Com isso, estende-se uma via de mão-dupla entre vida e imaginário, de modo que uma constitui e é constituída pelo outro. Como afirma Carlos Santeiro, personagem de Um romance de geração, trata-se das “possibilidades possíveis” do texto para além do texto, uma vez que o acordo entre narrador romanesco e leitor – e que sustenta o acontecimento literário – carrega muito dos pressupostos das diversas instâncias discursivas entre indivíduos, ficcionais ou não.
Costos y rendimientos de producción de tres néctares de manzana (pyrusmalusl) variedades Anna, Pensilvania y Winter

OpenAIRE

Darío Alberto Pinto Medina; Yesenia Fernández Vargas; Efrain Martinez Quintero

2016-01-01

La agroindustria en Colombia se muestra como una alternativa para dar aprovechamiento eficiente a materias primas de origen agropecuario, especialmente aquellas que por parámetros de calidad, no salen al mercado y generan cuantiosas pérdidas a agricultores y ganaderos; esta investigación se basó principalmente en el aprovechamiento de la manzana (Pyrus malus L), se elaboró néctar con tres variedades producidas en el departamento de Boyacá, Colombia: Pensilvania, Anna y Winter y se evaluaron l...
Education and research at the VR-1 Vrabec training reactor facility

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

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

International Nuclear Information System (INIS)

Petit, A.; Brie, M.

1996-01-01

The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs
Sandia reactor kinetics codes: SAK and PK1D

International Nuclear Information System (INIS)

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

1978-01-01

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

International Nuclear Information System (INIS)

Adam, E.R.

1997-01-01

As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)
Steam-generator replacement sets new marks

International Nuclear Information System (INIS)

Beck, R.L.

1995-01-01

This article describes how, in one of the most successful steam-generator replacement experiences at PWRs worldwide, the V C Summer retrofit exceeded plant goals for critical-path duration, radiation, exposure, and radwaste generation. Intensive planning and teamwork, combined with the firm support of station management and the use of mockups to prepare the work crews for activity in a radiological environment, were key factors in the record performance achieved by South Carolina Electric and Gas Co (SCE and G) in replacing three steam generators at V C Summer nuclear station. The 97-day, two-hour breaker-to-breaker replacement outage -- including an eight-day delay for repair of leak in a small-bore seal-injection line of a reactor coolant pump (unrelated to the replacement activities) -- surpassed the project goal by over one day. Moreover, the outage was only 13 hours shy of the world record held by Virginia Power Co's North Anna Unit 1
Claiming justice: knowing mental illness in the public art of Anna Schuleit's 'Habeas Corpus' and 'Bloom'.

Science.gov (United States)

Bell, Susan E

2011-05-01

This study investigates two public art performances by artist Anna Schuleit in the early 2000s commemorating the life and history of two state hospitals ('asylums') in Massachusetts and the people who built, worked, and were patients in them. Public art is made for and sited in the public domain, outside, freely accessible, frequently collaborative, and often ephemeral. This study addresses a series of questions: What can public art 'do' for understanding mental illness? What use is a public art project for those living with (and caring for those who live with) mental illness? How can a public work of art sustain and portray meaning in an expressive way, open up a shared discursive space, and demand witness through embodiment?
Ageing problems and renovation programme of ET-RR-1 reactor

International Nuclear Information System (INIS)

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

1995-01-01

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

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

African Journals Online (AJOL)

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

International Nuclear Information System (INIS)

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

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)
RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

International Nuclear Information System (INIS)

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

1987-01-01

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

Directory of Open Access Journals (Sweden)

Karin Birge Gilardoni-Büch

2014-04-01

Both of them succeed in emigrating and surviving the destruction of the European Jews. Anna Seghers comes back in 1947 to the Soviet occupied zone and will later hold significant offices inside East Germany's cultural life. Mascha Kaléko will however come back only now and then to the German Federal Republic. Did their writing change after the Shoah? And if so, to what extent? Moreover, do the texts mirror the situation in which their authors found themselves? Aren’t they linked to a specific social condition, that of a woman? To what extent do specific female and Jewish experiences take shape in their works?
A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

International Nuclear Information System (INIS)

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

2015-01-01

In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe
A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe
Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

International Nuclear Information System (INIS)

Kuruc, J.; Matel, L.

2007-01-01

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

International Nuclear Information System (INIS)

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

2003-01-01

FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de
Evaporation rate measurement in the pool of IEAR-1 reactor

International Nuclear Information System (INIS)

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

2000-01-01

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)
Analysis of COMS-1 North-South Station Keeping Method

Directory of Open Access Journals (Sweden)

Hae-Yeon Kim

2005-12-01

Full Text Available The perturbations caused by the Sun and the Moon are predominantly out-of-plane effects causing a change in the inclination and in the right ascension of ascending node of a geostationary satellite. Due to the change of the inclination, subsatellite latitude of the geostationary satellite has a daily variations of the same magnitude of the inclination. Therefore we need a facility to control the orbital inclination and right ascension of ascending node for maintaining the satellite position in specified subsatellite latitude boundary using thrusters. In this paper we studied North-South station keeping strategies of the COMS-1 such as Track-Back Chord Target (TBCT method, Maximum Compensation Target (MCT method and Minimum Fuel Target (MFT method. We accomplished those North-South station keeping maneuvers for one year starting from December 2008. The required velocity increments to maintain the satellite are estimated as MCT 52.6065m/s, TBCT 52.2383m/s, MFT 51.5428m/s, respectively. We demonstrated that TBCT and MFT methods are proper to North-South station keeping for COMS-1. MFT method showed the minimum required velocity increments whereas TBCT traced narrow inclination boundary area for North-South station keeping.
IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)
Biblioteca escolar das escolas reunidas Sant’anna do Paranahyba/MT (1936-1945: contribuições para o estudo de sua história.

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Rosimar Pires Alves

2016-12-01

Full Text Available Neste texto apresentam-se os resultados de pesquisa de mestrado vinculada ao Programa de Pós-Graduação em Educação da Universidade Estadual de Mato Grosso do Sul (UEMS, Unidade Universitária de Paranaíba, sobre o tema história das bibliotecas escolares na escola primária urbana em Paranaíba/MT. A pesquisa é oriunda dos resultados do projeto “Memória da Escola Primária em Paranaíba/MS (1946-1971”, desenvolvido no âmbito do Grupo de Estudos e Pesquisas em História e Historiografia da Educação Brasileira (GEPHEB. No projeto verificou-se em alguns documentos escolares localizados, reunidos e selecionados sobre as Escolas Reunidas Sant‟Anna do Paranahyba – um dos primeiros núcleos de escolarização do município de Paranaíba –, a fundação e a organização de uma biblioteca escolar a partir do ano de 1936, o que levou à constituição do objeto de estudo. Com isso traçou-se como objetivo contribuir para a produção de uma história das bibliotecas escolares em Mato Grosso do Sul e no Brasil, a partir da história da primeira biblioteca escolar organizada nas Escolas Reunidas Sant‟Anna do Paranahyba, mediante localização, reunião, seleção, organização e análise de fontes documentais sobre a implantação, o acervo, as especificidades e os sujeitos envolvidos. O recorte temporal abrange o período de 1936 – início das atividades da biblioteca escolar – até 1945, quando as Escolas Reunidas Sant‟Anna do Paranahyba foram transformadas no Grupo Escolar José Garcia Leal. A pesquisa fundamentou-se na abordagem histórica, no campo da educação, na vertente da Nova História que apresenta um leque de possibilidades no que se refere à pesquisa histórica em educação, constituindo um campo produtivo de estudos. Assim, a pesquisa realizada permitiu compreender que a Biblioteca Escolar das Escolas Reunidas Sant‟Anna do Paranahyba não se mostrou aquém do movimento em prol da leitura e das bibliotecas
A Brief Assessment of North Korea's Capacities for Building an Experimental LWR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Hyu; An, Jin Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2011-10-15

On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor
The North Korean nuclear dilemma

International Nuclear Information System (INIS)

Hecker, Siegfried S.

2004-01-01

The current nuclear crisis, the second one in ten years, erupted when North Korea expelled international nuclear inspectors in December 2002, then withdrew from the Nuclear Nonproliferation Treaty (NPT), and claimed to be building more nuclear weapons with the plutonium extracted from the spent fuel rods heretofore stored under international inspection. These actions were triggered by a disagreement over U.S. assertions that North Korea had violated the Agreed Framework (which froze the plutonium path to nuclear weapons to end the first crisis in 1994) by clandestinely developing uranium enrichment capabilities providing an alternative path to nuclear weapons. With Stanford University Professor John Lewis and three other Americans, I was allowed to visit the Yongbyon Nuclear Center on Jan. 8, 2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. On the basis of our visit, we were not able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. However
RA reactor operation and maintenance in 1992, Part 1

International Nuclear Information System (INIS)

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

1992-01-01

During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr
Public participation in energy related decision making: Six case studies. Final report

Energy Technology Data Exchange (ETDEWEB)

Clemente, F.; Cole, J.; Kloman, E.; McCabe, J.; Sawicki, P.

1977-12-01

Each of the six case studies documents public participation in Federal and/or state governmental decisions related to energy facility siting. Four of the cases involved decisions on specific facilities at specific sites, namely: (1) various state and federal licensing procedures for the Seabrook, New Hampshire nuclear facility; (2) the Maine Environmental Improvement Commission's denial of a permit for an oil refinery on Sears Island in Penobscot Bay; (3) the Atomic Energy Commission's amendment to the license for the Big Rock Point, Michigan, nuclear reactor to allow an increased level of plutonium-enriched fuel use; (4) the AEC's review, arising from disclosure of a geological fault, of the North Anna River, Virginia, nuclear facility. A fifth case documents a series of public meetings conducted in Pennsylvania by the Governor's Energy Council to consider the energy park concept. The sixth study was a narrative history and analysis of RM-50-1, a rulemaking proceeding conducted by the AEC in 1972 and 73 on emergency core cooling system operating standards.
Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

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

1999-01-01

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

International Nuclear Information System (INIS)

Salmenhaara, S.E.J.

2001-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)
Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

Qarṭaŷanna y Bāguh, cecas almohades, y la hipótesis de las

Directory of Open Access Journals (Sweden)

Vega Martín, Miguel

2006-06-01

Full Text Available
A huge hoard of Almohad (Muwaḥḥid coins was found in Priego de Cordoba (Cordoba, Spain in 1959. It contains several thousands of silver coins and is kept in the Archeological and Ethnological Museum of Cordoba. The hoard provides archeological evidences that Qarṭaŷanna and Bāguh, so far unregistered in Almohad coins, must be added to the list of Andalusi mints of that period. After identifying the two names of towns as Cartagena (Murcia, Spain and Priego de Cordoba, the hypothesis that the Almohad authorities issued series of coins as commemoration of relevant political or religious events is maintained.

En el Museo Arqueológico y Etnológico de Córdoba se conserva un copioso tesorillo de monedas almohades de plata compuesto por varios millares de piezas, hallado en Priego de Córdoba en 1959. Algunas de ellas permiten la lectura de dos nuevas cecas: las de Qarṭaŷanna y Bāguh, que pueden identificarse con las localidades españolas actuales de Cartagena y Priego. La multiplicidad de cecas almohades recibe aquí una explicación hipotética: se trataría de acuñaciones conmemorativas de acontecimientos políticos o religiosos.
EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

Energy Technology Data Exchange (ETDEWEB)

Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shimskey, R. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, N. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, R. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-15

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.
Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

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

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt
Computer-aided testing and operational aids for PARR-1 nuclear reactor

International Nuclear Information System (INIS)

Ansari, S.A.

1990-01-01

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Directory of Open Access Journals (Sweden)

Anna Zawadzka

2012-12-01

Full Text Available On subjects that cannot be put to sleep. Anna Zawadzka in an interview with Jerzy Jedlicki The conversation with Jerzy Jedlicki is a part of series of interviews starting with the question about the subjects, that the interlocutors during his or her academical career were advised (by different persons and for different reasons not to pursue. Jedlicki mentions about three such issues: Poles and Jews during II World War period, Israel and Palestine, atheism and faith. He points out that nobody discouraged him from pursuing these subjects – it was himself, who refrained from being vocal on these matters. He clarifies why he used to censor himself, analyses how all the subjects work today in Polish public sphere and explains why he broke the silence several times. Tematy, których nie da się uśpić. Z Jerzym Jedlickim rozmawia Anna Zawadzka Rozmowa z Jerzym Jedlickim przynależy do cyklu wywiadów, które rozpoczynają się pytaniem o tematy, które w toku kariery naukowej były rozmówcy z różnych przyczyn i przez różne osoby odradzane. Jedlicki wymienia trzy takie kwestie: Polacy i Żydzi w okresie okupacji, Izrael i Palestyna oraz ateizm i wiara. Zaznacza, że nikt mu nie odradzał ich podejmowania, a jedynie on sam powstrzymywał się przed zabieraniem publicznie głosu w tych sprawach. Wyjaśnia dlaczego się autocenzurował, analizuje, w jaki sposób wszystkie te trzy tematy funkcjonują dziś w polskiej sferze publicznej oraz tłumaczy, dlaczego kilkakrotnie przerwał jednak milczenie.
Spent fuel management plans for the FiR 1 Reactor

International Nuclear Information System (INIS)

Salmenhaara, S. E. J.

2002-01-01

The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)
Experimental study of the IPR-R1 TRIGA reactor power channels responses

International Nuclear Information System (INIS)

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

2015-01-01

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

International Nuclear Information System (INIS)

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

1998-01-01

IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)
RA reactor operation and maintenance in 1996, Part 1

International Nuclear Information System (INIS)

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

1996-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr
Possible future roles for fast breeder reactors Part 1 and 2

International Nuclear Information System (INIS)

1978-06-01

Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks
The AMPS 1.5 MW low-pressure compact reactor

International Nuclear Information System (INIS)

Hewitt, J.S.

1987-01-01

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

International Nuclear Information System (INIS)

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

1977-07-01

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

International Nuclear Information System (INIS)

Hines, J. Wesley; Upadhyaya, Belle R.; Doster, J. Michael; Edwards, Robert M.; Lewis, Kenneth D.; Turinsky, Paul; Coble, Jamie

2011-01-01

on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The
Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

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

Science.gov (United States)

El-Genk, Mohamed S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.
Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

International Nuclear Information System (INIS)

El-genk, M.S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept
Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

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

1982-06-01

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

International Nuclear Information System (INIS)

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

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt
Estimate of radiocaesium derived FNPP1 accident in the North Pacific Ocean

Science.gov (United States)

Inomata, Yayoi; Aoyama, Michio; Tsubono, Takaki; Tsumune, Daisuke; Yamada, Masatoshi

2017-04-01

134Cs and 137Cs (radiocaesium) were released to the North Pacific Ocean by direct discharge and atmospheric deposition released from the TEPCO Fukushima Dai-ichi Nuclear Power Plant (FNPP1) accident in 2011. After the FNPP1 accident, measurements of 134Cs and 137Cs were conducted by many researches. However, those results are only snapshots in order to interpret the distribution and transport of the released radiocaesium on a basin scale. It is recognized that estimation of the total amount of released 134Cs and 137Cs is necessary to assess the radioecological impacts of their release on the environment. It was reported that the inventory of 134Cs or 137Cs on the North Pacific Ocean after the FNPP1 accident was 15.2-18.3 PBq based on the observations (Aoyama et al., 2016a), 15.3±1.6 PBq by OI analysis (Inomata et al., 2016), 16.1±1.64 PBq by global ocean model (Tsubono et al., 2016). These suggest that more than 75 % of the atmospheric-released radiocaesium (15.2-20.4 PBq; Aoyama et al., 2016a) were deposited on the North Pacific Ocean. The radiocaesium from the atmospheric fallout and direct discharge were expected to mixing as well as diluting near the coastal region and transported eastward across the North Pacific Ocean in the surface layer. Furthermore, radicaesium were rapidly mixed and penetrated into the subsurface water in the North Pacific Ocean in winter. It was revealed that these radiocaesium existed in the Subtropical Mode Water (STMW, Aoyama et al., 2016b; Kaeriyama et al., 2016) and Central Mode Water (CMW, Aoyama et al., 2016b), suggesting that mode water formation and subduction are efficient pathway for the transport of FNPP1 derived radiocaesium into the ocean interior within 1-year timescale. Kaeriyama et al. (2016) estimated the total amount of FNPP1 derived radiocaesium in the STMW was 4.2 ± 1.1 PBq in October-November 2012. However, there is no estimation of the amount of radiocaesium in the CMW. Therefore, it is impossible to discuss

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

Energy Technology Data Exchange (ETDEWEB)

Forlerer, Elena; Palacios, Tulio A [comps.

1998-07-01

After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation.
Benchmark testing of Canadol-2.1 for heavy water reactor

International Nuclear Information System (INIS)

Liu Ping

1999-01-01

The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5
Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.
Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)
Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

Energy Technology Data Exchange (ETDEWEB)

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

1999-08-01

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

International Nuclear Information System (INIS)

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

1999-01-01

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

International Nuclear Information System (INIS)

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

1996-01-01

This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)
Modifications done in the IPR-R1 reactor and their auxiliary systems

International Nuclear Information System (INIS)

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

1986-01-01

The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt
Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities
Safety evaluation report related to the operation of Catawba Nuclear Station, Units 1 and 2. Docket Nos. 50-413 and 50-414. Suppl. 1

International Nuclear Information System (INIS)

1983-04-01

This reort supplements the Safety Evaluation Report (NUREG-0954) issued in February 1983 by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission with respect to the application filed by Duke Power Company, North Carolina Municipal Power Agency Number 1, North Carolina Membership Corporation, and Saluda River Electric Cooperative, Inc. as applicants and owners, for licenses to operate the Catawba Nuclear Station, Units 1 and 2 (Docket Nos. 50-413 and 50-414, respectively). The facility is located in York County, South Carolina, approximately 9.6 km (6 mi) north of Rock Hill and adjacent to Lake Wylie. This supplement provides more recent information regarding resolution or updating of some of the open and confirmatory items and license conditions identified in the Safety Evaluation Report, and discusses the recommendations of the Advisory Committee on Reactor Safeguards in its report dated March 15, 1983
Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)
Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

International Nuclear Information System (INIS)

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

1990-01-01

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

International Nuclear Information System (INIS)

Bars, Bruno; Kall, Leif

1982-01-01

The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)
Dose measurements in controlled area of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

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

2005-01-01

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-06-01

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

International Nuclear Information System (INIS)

Kolros, Antonin; Huml, Ondrej; Kos, Josef

2008-01-01

Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity
Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

International Nuclear Information System (INIS)

Olsson, S.; Andermo, L.

1980-01-01

This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)
RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

International Nuclear Information System (INIS)

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

1986-01-01

In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr
Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

Energy Technology Data Exchange (ETDEWEB)

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

2006-01-01

The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.
RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

International Nuclear Information System (INIS)

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

1988-01-01

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

IPR-R1 TRIGA research reactor decommissioning plan

International Nuclear Information System (INIS)

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

2008-01-01

The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)
The Chernobyl reactor accident. Pt. 1 and 2

International Nuclear Information System (INIS)

1986-06-01

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

Science.gov (United States)

Clancey, Noel; Horney, Barbara; Burton, Shelley; Birkenheuer, Adam; McBurney, Scott; Tefft, Karen

2010-04-01

A 4-6-mo-old female red fox (Vulpes vulpes) was presented to the Atlantic Veterinary College (AVC) Teaching Hospital, Prince Edward Island, Canada. On presentation, the fox was weak and had pale mucous membranes. A complete blood count and a serum biochemistry profile were performed. Blood smear examination revealed low numbers of erythrocytes containing centrally to paracentrally located, single, rarely multiple, approximately 1 x 2 microm, oval to round organisms with morphology similar to Babesia microti. Polymerase chain reaction testing and DNA sequencing of the Babesia species 18S rRNA gene were performed on DNA extracted from whole blood. Results were positive for a Babesia microti-like parasite genetically identical to Babesia (Theileria) annae. The fox was euthanized due to poor prognosis for recovery. Necropsy examination revealed multifocal to locally extensive subacute nonsuppurative meningoencephalitis, an eosinophilic broncho-pneumonia, a moderate diffuse vacuolar hepatopathy, and lesions associated with blunt trauma to the left abdominal region. This is the first reported case of a red fox in Canada infected with a piroplasm. It remains uncertain whether the presence of this hemoparasite in this fox was pathogenic or an incidental finding. The potential for competent vectors of Babesia species on Prince Edward Island, the potential for this Babesia microti-like parasite to infect other wild and domestic canids, and the significance of this parasite to the health of infected individuals are yet to be determined.
Current activities at the FiR 1 TRIGA reactor

International Nuclear Information System (INIS)

Salmenhaara, Seppo

2002-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover
Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

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

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)
Expanding the storage capability at ET-RR-1 research reactor at Inshass

International Nuclear Information System (INIS)

Sultan, Mariy M.; Khattab, M.

1999-01-01

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

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1985-01-01

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

International Nuclear Information System (INIS)

1988-03-01

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

NARCIS (Netherlands)

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

Biotransformation of 1,1,1-trichloroethane (CH3CCl3) was observed in an anaerobic packed-bed reactor under conditions of both sulfate reduction and methanogenesis. Acetate (1 mM) served as an electron donor. CH3CCl3 was completely converted up to the highest investigated concentration of 10 mu M.
Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

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

2009-01-01

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring
Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1985-01-01

This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr
Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

International Nuclear Information System (INIS)

Andermo, L.; Sundman, B.

1974-04-01

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

Energy Technology Data Exchange (ETDEWEB)

N/A

1999-05-14

services. The procurement process included the environmental review specified in DOE's NEPA regulations in 10 CFR 1021.216. The six reactors selected are Catawba Nuclear Station Units 1 and 2 in South Carolina McGuire Nuclear Station Units 1 and 2 in North Carolina, and North Anna Power Station Units 1 and 2 in Virginia. The Supplement describes the potential environmental impacts of using MOX fuel in these six specific reactors named in the DCS proposal as well as other program changes made since the SPD Draft EIS was published.
Supplement to the Surplus Plutonium Disposition Draft Environmental Impact Statement

International Nuclear Information System (INIS)

1999-01-01

included the environmental review specified in DOE's NEPA regulations in 10 CFR 1021.216. The six reactors selected are Catawba Nuclear Station Units 1 and 2 in South Carolina McGuire Nuclear Station Units 1 and 2 in North Carolina, and North Anna Power Station Units 1 and 2 in Virginia. The Supplement describes the potential environmental impacts of using MOX fuel in these six specific reactors named in the DCS proposal as well as other program changes made since the SPD Draft EIS was published
New digital control system for the operation of the Colombian research reactor IAN-R1

International Nuclear Information System (INIS)

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

2015-09-01

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

International Nuclear Information System (INIS)

Kirchner, W.L.; Meier, K.L.

1984-01-01

A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, ''walk-away safe'' design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (OandM) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System
Low power unattended defense reactor

International Nuclear Information System (INIS)

Kirchner, W.L.; Meier, K.L.

1984-01-01

A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System
Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias

2005-01-01

The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical
Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

International Nuclear Information System (INIS)

Andermo, L.

1977-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-03-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.
Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

International Nuclear Information System (INIS)

El-Messeiry, A.M.

1996-01-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs
COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

International Nuclear Information System (INIS)

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

1964-01-01

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-12-15

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)
Welding of the A1 reactor pressure vessel

International Nuclear Information System (INIS)

Becka, J.

1975-01-01

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

Science.gov (United States)

Ventegodt, Søren; Clausen, Birgitte; Merrick, Joav

2006-03-07

When we experience life events with overwhelming emotional pain, we can escape this pain by making decisions (in our mind) that transfer responsibility from our existence to the surrounding world. By doing this, we slowly destroy the essence of our being, health, quality of life, and ability to function. The case of Anna is an excellent example of such a systematic destruction of self, done to survive the extreme pressure from childhood abuse and sexual abuse. The case study shows that the damage done to us by traumatic events is not on our body or soul, but rather our philosophy of life. The important consequence is that we can heal our existence by letting go of the negative decisions taken in the past painful and traumatic situations. By letting go of the life-denying sentences, we come back to life and take responsibility for our own life and existence. The healing of Anna's existence was done by existential holistic therapy. Although the processing did not always run smoothly, as she projected very charged material on the therapists on several occasions, the process resulted in full health and a good quality of life due to her own will to recover and heal completely. The case illustrates the inner logic and complexity of intensive holistic therapy at the most difficult moment, where only a combination of intensive medical, psychiatric, and sexological treatment could set her free. In the paper, we also present a meta-perspective on intensive holistic therapy and its most characteristic phases.
Determination flux in the Reactor JEN-1

International Nuclear Information System (INIS)

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

1960-01-01

This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs
Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

International Nuclear Information System (INIS)

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

1995-06-01

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

Science.gov (United States)

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

2017-11-01

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

International Nuclear Information System (INIS)

Pruett, D.P.; Richards, W.J.; Heidel, C.C.

1982-01-01

Neutron radiography of the fuel undergoing examination in the argon cell is performed in the NRAD Facility and is one of many examinations performed on the fuel. The reactor and examination procedure are described. The new radiography system, developed to expand the present radiography capabilities to radiograph both irradiated and unirradiated specimens and to provide for the development of new radiography techniques without interfering with the argon cell production schedule is presented
IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

West, C D [comp.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.
IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

West, C.D.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface
Energy balance of algal biomass production in a 1-ha “Green Wall Panel” plant: How to produce algal biomass in a closed reactor achieving a high Net Energy Ratio

International Nuclear Information System (INIS)

Tredici, M.R.; Bassi, N.; Prussi, M.; Biondi, N.; Rodolfi, L.; Chini Zittelli, G.; Sampietro, G.

2015-01-01

Highlights: • Tetraselmis suecica production in a 1-ha GWP plant in Tuscany (Italy) has a NER < 1. • Major energy costs are embodied energy of GWP and mixing. • In a suitable location (North Africa) the NER increases by 40%. • Integration of photovoltaic in the GWP allows to achieve a NER of 1.7. • T. suecica cultivated in a GWP plant can yield up to 30 t of protein ha −1 year −1 . - Abstract: The annual productivity of Tetraselmis suecica in a 1-ha Green Wall Panel-II (GWP-II) plant in Tuscany (Italy) is 36 t (dry weight) ha −1 year −1 , which corresponds to an energy output of 799 GJ ha −1 year −1 . The energy inputs necessary to attain that productivity amount to 1362 GJ ha −1 year −1 , mainly given by the embodied energy of the reactor (about 30%), mixing (about 40%), fertilizers (11%) and harvesting (10%). The Net Energy Ratio (NER) of T. suecica production is thus 0.6. In a more suitable location (North Africa) productivity nearly doubles, reaching 66 t ha −1 year −1 , but the NER increases only by 40% and the gain (difference between output and inputs) remains negative. In a GWP-II integrated with photovoltaics (PV), the NER becomes 1.7 and the gain surpasses 600 GJ ha −1 year −1 . Marine microalgae cultivation in a GWP plant, in a suitable location, can attain high biomass productivities and protein yields 30 times higher than those achievable with traditional crops (soya). When the GWP reactor is integrated with PV, the process attains a positive energy balance, which substantially enhances its sustainability
Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

International Nuclear Information System (INIS)

1987-01-01

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

International Nuclear Information System (INIS)

Tikhomirov, L.N.

1995-01-01

The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs
RA reactor operation and maintenance in 1994, Part 1

International Nuclear Information System (INIS)

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

1994-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr
48 CFR 27.204-1 - Use of patented technology under the North American Free Trade Agreement.

Science.gov (United States)

2010-10-01

... under the North American Free Trade Agreement. 27.204-1 Section 27.204-1 Federal Acquisition Regulations... Patents and Copyrights 27.204-1 Use of patented technology under the North American Free Trade Agreement... the patent holder is from a country that is a party to the North American Free Trade Agreement (NAFTA...
Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

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

International Nuclear Information System (INIS)

Odoi, H.C.

2014-06-01

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

Fast breeder reactor at Kalkar. Pt. 2

International Nuclear Information System (INIS)

Degen, G.

1979-02-01

After a brief description of the previous development of the case the legal decisions are documented and commented on. The concept of the then FDP-Minister of Economy of North Rhine Westphalia (Riemer, Pu-combustion plant) is presented and the prospects and risk for the fast breeder reactor after the 3. partial construction license are discussed. (orig./HP) [de
New burnup calculation of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

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

2015-01-01

The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

International Nuclear Information System (INIS)

1986-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)
Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

International Nuclear Information System (INIS)

1978-05-01

An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization
Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

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

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal
75 FR 76495 - Virginia Electric and Power Company North Anna Power Station, Unit Nos. 1 and 2 Surry Power...

Science.gov (United States)

2010-12-08

..., Supplement 1 and 2, ``Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency.... Environmental Impacts of the Proposed Action The NRC has completed its environmental assessment of the proposed... NRC concludes that there are no significant environmental impacts associated with the proposed action...
Problems of nuclear reactor safety. Vol. 1

International Nuclear Information System (INIS)

Shal'nov, A.V.

1995-01-01

Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed
Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

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

International Nuclear Information System (INIS)

Coura, J.G.

1986-01-01

This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt
Report on the 9th workshop on the innovative water reactor for flexible fuel cycle

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

2006-07-01

The research on Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been performed in JAEA for the development of future innovative reactor. The workshop on FLWRs has been held every year since 1998 aiming at information exchange with other organizations such as universities, laboratories, utilities and vendors. The 9th workshop was held on March 1, 2006 under the joint auspices of JAEA and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 64 participants. The workshop began with presentation entitled 'Activities on Nuclear Science and Engineering Research and Collaboration with Industry in JAEA', followed by presentations entitled 'Progress of Research and Development on FLWR' and 'On Final Report of Feasibility Study (phase 2) on Commercialized FBR Cycle Systems'. Then two lectures followed: 'Core and Fuel Design on Super Light Water Reactor' by Tokyo University and 'Recent trends on the Development of Next Generation Nuclear Reactor' by Institute of Applied Energy. This report summarizes the lectures of the workshop. (author)
OSCAR-4 Code System Application to the SAFARI-1 Reactor

International Nuclear Information System (INIS)

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

2008-01-01

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

International Nuclear Information System (INIS)

Novelli, A.

1982-01-01

The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)
Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

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

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)
Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

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

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.
SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1973-07-01

By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)
Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

International Nuclear Information System (INIS)

1987-08-01

Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)
Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

International Nuclear Information System (INIS)

1982-01-01

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

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

International Nuclear Information System (INIS)

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

1976-01-01

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

International Nuclear Information System (INIS)

Karpeta, C.; Volf, K.

1974-01-01

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

International Nuclear Information System (INIS)

Hawari, Ayman I.; Moxom, Jeremy; Hathaway, Alfred G.; Brown, Benjamin; Gidley, David W.; Vallery, Richard; Xu, Jun

2009-01-01

An intense slow positron beam is in its early stages of operation at the 1-MW open-pool PULSTAR research reactor at North Carolina State University. The positron beam line is installed in a beam port that has a 30-cmx30-cm cross sectional view of the core. The positrons are created in a tungsten converter/moderator by pair-production using gamma rays produced in the reactor core and by neutron capture reactions in cadmium cladding surrounding the tungsten. Upon moderation, slow (∼3 eV) positrons that are emitted from the moderator are electrostatically extracted, focused and magnetically guided until they exit the reactor biological shield with 1-keV energy, approximately 3-cm beam diameter and an intensity exceeding 6x10 8 positrons per second. A magnetic beam switch and transport system has been installed and tested that directs the beam into one of two spectrometers. The spectrometers are designed to implement state-of-the-art PALS and DBS techniques to perform positron and positronium annihilation studies of nanophases in matter.
RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

International Nuclear Information System (INIS)

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

2002-01-01

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)
Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

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

International Nuclear Information System (INIS)

Zhang Min; Peng Muzhang; Ding Lequn

1991-01-01

If a district heating reactor is set up in north seashore cities of China and operated in such a way that it supplies heating for inhabitants in winter and produces fresh water from seawater in summer, the load factor of the reactor would be increased, and the problem of fresh water shortage would be solved, in addition, it would benefit environment and promote tourism
Research reactor core conversion guidebook. V.1: Summary

International Nuclear Information System (INIS)

1992-04-01

In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs
Reed Reactor Facility final report, September 1, 1995--August 31, 1996

International Nuclear Information System (INIS)

1997-01-01

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.
NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

International Nuclear Information System (INIS)

Paiva, Rosemeire P.; Salvetti, Tereza C.

2005-01-01

Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)
Life extension of the St. Lucie unit 1 reactor vessel

International Nuclear Information System (INIS)

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

1991-01-01

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

International Nuclear Information System (INIS)

Ball, G.

2003-01-01

Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)
Shielding assessment of the ETRR-1 Reactor Under power upgrading

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

International Nuclear Information System (INIS)

Chambadal, P.; Pascal, M.

1955-01-01

The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr
Development of a training simulator to operators of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carvalho, Ricardo Pinto de

2006-01-01

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

International Nuclear Information System (INIS)

Bucheit, D.M.; Smith, N.A.

1987-01-01

Virginia Power has been performing in-house reload core design and safety analysis for several years. These analyses have been in support of North Anna units 1 and 2 and Surry units 1 and 2, all of which are three-loop pressurized water reactor plants designed and built by Westinghouse. Historically, Virginia Power first developed the capability to design and optimize its own core loading patterns in the early 1970's. This development effort was driven by the need to establish in-house control of the fuel management process, thereby ensuring that energy generation requirements are met in an economically optimum fashion. It soon became obvious that reload design and safety analysis processes are so integrally coupled that in order to perform the fuel management function in an effective manner, in-house capability in both areas needed to be developed. After reviewing the spectrum of economic, safety and operational constraints which affect the reload design and analysis process, an integrated model of the process is presented in flow chart format. This is followed by several specific examples which illustrate the interplay between sound fuel management practice and the assurance of plant safety using in-house analysis techniques
An overview of the RECH-1 reactor conversion

International Nuclear Information System (INIS)

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

2000-01-01

The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)
Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)
Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

International Nuclear Information System (INIS)

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

2011-01-01

JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)
An economic analysis of stretch-out for Angra-1 reactor

International Nuclear Information System (INIS)

Sakai, M.

1989-01-01

An application of NUCOST code for calculating nuclear energy cost is presented. Ann optimization of stretch-out for Angra-1 reactor based on international costs of nuclear fuel, operation and maintenance is done. (M.C.K.)
2014 NCFMP Lidar: Statewide North Carolina (Phase 1)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Geographic Extent: North Carolina Area of Interest for Sandy, covering approximately 9,396 square miles. Dataset Description: The North Carolina - Sandy LiDAR...
North Carolina Statewide Lidar DEM 2014 Phase 1

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Geographic Extent: North Carolina Area of Interest for Sandy, covering approximately 9,396 square miles. Dataset Description: The North Carolina - Sandy LiDAR...
The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

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

International Nuclear Information System (INIS)

Lakatos, L.

1975-01-01

The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)
1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Kim, Jae Cheol

2007-02-01

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

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data
The North Sea field development guide. V.1: Northern North Sea. V.2: Southern North Sea. 6. ed.

International Nuclear Information System (INIS)

Anon.

1996-08-01

The 1997/8, sixth edition is the first to be divided into two volumes. Volume 1 covers the central and northern North Sea areas; volume 2 contains the southern North Sea as well as the Irish and German sectors. The pages are numbered consecutively over the two volumes, with page numbers greater than 702 contained in the second volume. There are three index sections. Main index. Arranged by national sector (UK, Norway etc.) Within each sector the entries are alphabetical by operator name. This index contains page numbers for the book entries: the other two index sections should be used with the main index to find the exact location of an entry; Index by field. If the reader knows a field name (e.g. Kittiwake) but not the operator or the national sector, this index will reference them; Index by installation. Each installation (''Fulmar SALM'', ''Togi'') is named, giving the operator, field and national sector where it is located. This index is also useful for locating particular kinds of installations, such as subsea completions; The book is intended to provide a factual overview of field development activity in the North Sea (a term loosely used to include the Irish Sea and the Baltic Sea). The aim is therefore to provide some background, specifications and history on every offshore installation in that area. Speculative or evaluative commentary is avoided where possible. No attempt has been made to forecast the form or probability of future developments, except in those few instances where announcements have been made by the oil company itself, and these are, clearly indicated. (UK)
Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)
. Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

International Nuclear Information System (INIS)

Yusuf, I; Mati, A. A.

2010-01-01

The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.
VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

International Nuclear Information System (INIS)

Matejka, K.; Sklenka, L.

2003-01-01

The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)
FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

International Nuclear Information System (INIS)

Auterinen, I.; Salmenhaara, S.E.J. . Author

2004-01-01

The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)
Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised
Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.
Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised
Nuclear material control at IEA-R1 nuclear research reactor

International Nuclear Information System (INIS)

1988-01-01

The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt
Experimental measurement of the refrigerant temperature of the TRIGA Mark III reactor of the ININ

International Nuclear Information System (INIS)

Gallardo S, L.F.; Alonso V, G.

1991-08-01

With the object of knowing the axial temperature profile of the refrigerant in the core of the TRIGA Mark III reactor of the ININ, the temperatures of this, at the enter, in the center and the exit of the core were measured, in the positions: west 2, north 2 and south 1. This was made by means of the thermo pars introduction mounted in aluminum guides, connected to a measurer of digital temperature, whose resolution is of ± 0.1 C. The measurements showed a bigger heating of the refrigerant in the superior half of the core, that which suggests that the axial profile of temperature of the reactor is not symmetrical with respect to the center or that those temperature measurements in the center are not correct. (Author)
Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

Energy Technology Data Exchange (ETDEWEB)

Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

1999-12-31

Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

Reed Reactor Facility final report, September 1, 1994--August 31, 1995

International Nuclear Information System (INIS)

1997-01-01

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits
Reed Reactor Facility final report, September 1, 1994--August 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.
Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

International Nuclear Information System (INIS)

Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

2010-01-01

In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)
Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

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

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.
1DB, a one-dimensional diffusion code for nuclear reactor analysis

International Nuclear Information System (INIS)

Little, W.W. Jr.

1991-09-01

1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs
Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-08-15

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.
Dominant accident sequences in Oconee-1 pressurized water reactor

International Nuclear Information System (INIS)

Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

1985-04-01

A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling
NCSU PULSTAR reactor instrumentation upgrade. Final technical report, September 6, 1990--March 19, 1993

International Nuclear Information System (INIS)

Bilyj, S.J.; Perez, P.B.

1993-01-01

The Nuclear Reactor Program at North Carolina State University initiated an upgrade program at the NCSU PULSTAR Reactor in 1990. Twenty-year-old instrumentation is currently undergoing replacement with solid-state and current technology equipment. The financial assistance from the United States Department of Energy has been the primary source of support. This report provides the status of the first two phases of the upgrade program
RA reactor operation and maintenance in 1989, Part 1

International Nuclear Information System (INIS)

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

1989-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr
Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators
System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

International Nuclear Information System (INIS)

Ellis, R.J.

2000-01-01

The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors
Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Directory of Open Access Journals (Sweden)

Montserrat Palau

2015-01-01

Full Text Available Spain suffered under the climate of state terrorism from the dictatorship of 1939 until the 1970s, and during the period including the last years of General Franco’s life to the first years of the transition to democracy. Within this historical context and during the same period, Catalonia suffered more under state terrorism than under the violence of other armed groups. The result of this was that the literature on this subject is either minimal or was written much later. Moreover, the focus on women terrorists was also scarce. Even at present, in situations of armed conflict, women are still seen as victims or as booty – not as active participants in violence. It is the distance in time and context that marks the narrative discourse of La terrorista desarmada (The Unarmed Terrorist (2002 by Anna Vilaseca. The story is set at the end of the Franco regime and illustrates the pervading binaries defining women as submissive and peace-loving.
Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1988-01-01

Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt
Core calculations for the upgrading of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

1998-01-01

The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)
Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.
North Korea's nuclear weapons development. Implications for future policy

International Nuclear Information System (INIS)

Pollack, J.D.

2010-01-01

denuclearization. But Pyongyang further asserts that it will retain its nuclear weapons 'until the denuclearization of the Korean Peninsula and the world is realized'. The North's problematic involvement in nuclear commerce and technology transfer adds significantly to this highly perturbing picture. The operative policy issues for the United States therefore focus less on an early end to the program, and more on how to mitigate the potential security risks resulting from North Korean behavior and strategic intentions.U.S. views of North Korean nuclear weapon development and the negotiability of this issue have shifted significantly over the years. Three recent developments in particular have diminished political support for accommodation and engagement: (1) the North's participation in the construction of a nuclear reactor in Syria destroyed in an Israeli attack in September 2007; (2) North Korea's second nuclear test of May 2009 and its avowed claims to status as a nuclear weapons state; and (3) the far more cautionary attitudes of South Korean President Lee Myung-bak, compared to the policies of his two immediate predecessors, Kim Dae-jung and Roh Moo-hyun, both of whom were deeply committed to engagement with the North. Despite some partial successes in restraining the DPRK's nuclear development (notably, the partial disablement of its graphite moderated reactor and related facilities in 2008), these accomplishments have proven reversible, ephemeral, or both. This pattern is both dismaying and worrisome. For two and half decades, North Korea has stymied, delayed, circumvented, and otherwise foiled the efforts of allies, adversaries, and the IAEA to ensure the DPRK's compliance with its non-proliferation obligations, all the while retaining the engineering and industrial capabilities necessary for nuclear weapons development. North Korea has also engaged in unambiguous violations of its declared commitments to denuclearization. In January 2003, the DPRK withdrew from the NPT
Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

2005-01-01

The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)
Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

Energy Technology Data Exchange (ETDEWEB)

Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

2014-08-15

IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.
Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Operating reactors licensing actions summary. Volume 5, Number 1

International Nuclear Information System (INIS)

1985-03-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program
Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

International Nuclear Information System (INIS)

Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

2008-01-01

The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation
Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

Energy Technology Data Exchange (ETDEWEB)

Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

2008-10-15

The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.
Source term determination from subcritical multiplication measurements at Koral-1 reactor

International Nuclear Information System (INIS)

Blazquez, J.B.; Barrado, J.M.

1978-01-01

By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)
Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

International Nuclear Information System (INIS)

Axtmann, R.C.; Perkins, H.K.

1977-08-01

Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems
Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

Directory of Open Access Journals (Sweden)

Muhammad Atta

2011-01-01

Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.
Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)
Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)
Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

International Nuclear Information System (INIS)

D'Utra Bitelli, U.

1993-01-01

This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)
Consequence analysis for nuclear reactors, Yongbyon

International Nuclear Information System (INIS)

Kang, Taewook; Jae, Moosung

2017-01-01

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea. (author)
AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

International Nuclear Information System (INIS)

Hansen, W.

2003-01-01

Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de
Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

International Nuclear Information System (INIS)

Balogun, G.I.; Jonah, S.A.; Umar, I.M.

2005-01-01

Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)
Two Austro-Hungarian Women Writers, Anna Tutsek and Terka Lux, Creating New Urban Identities in Early Twentieth-Century Budapest

Directory of Open Access Journals (Sweden)

Judit Kádár

2016-01-01

Full Text Available In this paper, I examine some literary texts of two turn-of-the century Hungarian women writers, Anna Tutsek and Terka Lux who left behind their childhood environment in remote regions of the Austro-Hungarian Monarchy in order to move to Budapest, the capital of the eastern part of the Empire. Assuming that individuals hold multiple identities that are flexible and inevitably affected by environmental and social changes, my main focus is on the transformation of their ethnic, regional, occupational and gender identity influenced by the disengagement from their birthplace. Within a context of Hungarian−Eastern-European women's social history, I investigate how migration had led them to reshape their original identities and create new ones and how these emigrant writers reacted to the loss of cultural and social norms in which they had previously lived.
Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

International Nuclear Information System (INIS)

1995-01-01

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period
Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-31

This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.
Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)
Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database
Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.
Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Adib, M.; Naguib, K.; Morcos, H.N

2001-09-01

The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.
Mitochondrial haplogroup H1 in north Africa: an early holocene arrival from Iberia.

Directory of Open Access Journals (Sweden)

Claudio Ottoni

Full Text Available The Tuareg of the Fezzan region (Libya are characterized by an extremely high frequency (61% of haplogroup H1, a mitochondrial DNA (mtDNA haplogroup that is common in all Western European populations. To define how and when H1 spread from Europe to North Africa up to the Central Sahara, in Fezzan, we investigated the complete mitochondrial genomes of eleven Libyan Tuareg belonging to H1. Coalescence time estimates suggest an arrival of the European H1 mtDNAs at about 8,000-9,000 years ago, while phylogenetic analyses reveal three novel H1 branches, termed H1v, H1w and H1x, which appear to be specific for North African populations, but whose frequencies can be extremely different even in relatively close Tuareg villages. Overall, these findings support the scenario of an arrival of haplogroup H1 in North Africa from Iberia at the beginning of the Holocene, as a consequence of the improvement in climate conditions after the Younger Dryas cold snap, followed by in situ formation of local H1 sub-haplogroups. This process of autochthonous differentiation continues in the Libyan Tuareg who, probably due to isolation and recent founder events, are characterized by village-specific maternal mtDNA lineages.

Calculation of the main neutron parameters of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Ojima, Mario Katsuhiko

1977-01-01

The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined
TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

Energy Technology Data Exchange (ETDEWEB)

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.
TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

International Nuclear Information System (INIS)

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead
Distribution of Hanford reactor produced radionuclides in the marine environment, 1961-73

International Nuclear Information System (INIS)

Seymour, A.H.

1980-01-01

At Hanford (U.S.A.), the plutonium-producing reactors were in operation during 1944-1971. The period of maximum reactor operation was 1955-1965, when eight reactors were in operation. The reactor deactivation programme began in 1965 and the last reactor was deactivated in 1971. All these reactors were cooled by Columbia River water which passed through the reactors and then was discharged to the river and ultimately to the North Pacific. The Laboratory of Radiation Ecology (LRE) of the University of Washington started an environmental survey programme in 1965 and continued it upto 1973 i.e. even after the last plutonium producing reactor was deactivated. The programme objectives were: (1) to find the geographical distribution and concentration of Hanford produced radionuclides in water, sediments and biota of the marine environment, (2) to relate the operation of the Hanford reactors during the period of deactivation to the concentration of radionuclides in marine organisms, and (3) to observe the rate at which the marine organisms cleansed themselves of 65 Zn after the primary source had been removed. An account of the programme and highlights of the observations are reported. Most of the radioactivity entering the river water and marine organisms was due to 51 Cr, 65 Zn and 32 P of which 65 Zn was found to be the most abundant radionuclide in the biological samples. The rate of radioactivity from the river water entering into the Ocean was about 1000 curies per day and it did not produce any observable effects on populations of marine organisms. The internal dose to man from 65 Zn via seafoods was only a small fraction of the permissible dose for individual members of the population. (M.G.B.)
Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

Energy Technology Data Exchange (ETDEWEB)

Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)
Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

International Nuclear Information System (INIS)

Kulcinski, G.L.; Maynard, C.W.

1984-04-01

Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors
Virginia Power thermal-hydraulics methods

International Nuclear Information System (INIS)

Anderson, R.C.; Basehore, K.L.; Harrell, J.R.

1987-01-01

Virginia Power's nuclear safety analysis group is responsible for the safety analysis of reload cores for the Surry and North Anna power stations, including the area of core thermal-hydraulics. Postulated accidents are evaluated for potential departure from nucleate boiling violations. In support of these tasks, Virginia Power has employed the COBRA code and the W-3 and WRB-1 DNB correlations. A statistical DNBR methodology has also been developed. The code, correlations and statistical methodology are discussed
Organization patterns of PWR power plants

International Nuclear Information System (INIS)

Leicman, J.

1980-01-01

Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)
PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

International Nuclear Information System (INIS)

Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

1990-06-01

Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs
Project management plan for the 105-C Reactor interim safe storage project. Revision 1

International Nuclear Information System (INIS)

Miller, R.L.

1997-01-01

In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)
Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

Energy Technology Data Exchange (ETDEWEB)

Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

1965-08-15

FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.
Neutron radiography in the IEA-R1 reactor

International Nuclear Information System (INIS)

Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

1988-08-01

Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt
TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

International Nuclear Information System (INIS)

Liles, D.R.; Mahaffy, J.H.

1986-07-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients
TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Liles, D.R.; Mahaffy, J.H.

1986-07-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.
Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

International Nuclear Information System (INIS)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

2014-10-01

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)
Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

2017-05-15

Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the
Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

International Nuclear Information System (INIS)

Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

2013-06-01

(Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate
Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

International Nuclear Information System (INIS)

Calvaresi, C.; Moreschi, L.F.

1983-01-01

PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)
Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

Energy Technology Data Exchange (ETDEWEB)

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

1982-01-01

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

Measurements of reactivity of reactor G1

International Nuclear Information System (INIS)

Bernot, J.; Koechlin, J.C.; Portes, L.; Teste du Bailler, A.

1957-01-01

The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [fr
Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

International Nuclear Information System (INIS)

Larry B. Wimmer

2001-01-01

The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000)
Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

International Nuclear Information System (INIS)

Moreira, Joao Manoel Losada; Frajndlich, Roberto

1995-01-01

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

Energy Technology Data Exchange (ETDEWEB)

Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

2017-08-30

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO₃-xPbTiO₃) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.
Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

International Nuclear Information System (INIS)

Tanabe, Fumiya

2011-01-01

In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)
Kilkanaście sekund. Z Konstantym Gebertem rozmawiają Joanna Tokarska-Bakir, Elżbieta Janicka i Anna Zawadzka

Directory of Open Access Journals (Sweden)

Joanna Tokarska-Bakir

2016-01-01

Full Text Available A few seconds. Konstanty Gebert in a conversation with Joanna Tokarska-Bakir, Elżbieta Janicka and Anna Zawadzka In 2012–2013, Poland was discussing ritual slaughter. Its adversaries claimed that the kind of animal slaughtering that makes the animal meat kosher and halal, causes extra suffering to animals and therefore should be banned. However, the postulate of banning ritual slaughter in Poland is not new and it goes back to the pre-war period. In the Studia Litteraria et Historica’s interview Konstanty Gebert talks about the history of Polish debate on the ritual slaughter, its anti-Semitic applications and about not anti-Semitic arguments risen in the discussion. He describes legal solutions for the ritual slaughter in other countries and addresses the issue of animal killing as a moral problem in Judaism. He also discusses the contemporary debate on the ritual slaughter: is it really in the name of the animal rights? Kilkanaście sekund. Z Konstantym Gebertem rozmawiają Joanna Tokarska-Bakir, Elżbieta Janicka i Anna Zawadzka W latach 2012–2013 przez Polskę przetoczyła się dyskusja o uboju rytualnym. Jego przeciwnicy argumentowali, że taki sposób zabijania zwierząt, by mięso z ich ciał było koszerne i halal, przysparza zwierzętom dodatkowych cierpień, dlatego powinien być zakazany. Jednak postulat, by zakazać uboju rytualnego, nie jest w Polsce niczym nowym. Jego tradycja sięga okresu przedwojennego. W wywiadzie dla czasopisma „Studia Litteraria et Historica” Konstanty Gebert opowiada m.in. o historii polskiej debaty na temat uboju rytualnego, o antysemickich z niej użytkach i o nieantysemickich argumentach w niej podnoszonych, o rozwiązaniach prawnych dotyczących uboju rytualnego w innych krajach, a także o zabijaniu zwierząt jako problemie moralnym w obrębie judaizmu. Zastanawia się także, czy współczesna debata o uboju rytualnym jest faktycznie w interesie zwierząt.
Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

DEFF Research Database (Denmark)

Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...
Reactor of the XXI century

International Nuclear Information System (INIS)

Zhotabaev, Zh.R.; Solov'ev, Yu.A.

2001-01-01

The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed
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
Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

International Nuclear Information System (INIS)

Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

2000-01-01

The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar
Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

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

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

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

International Nuclear Information System (INIS)

1969-05-01

Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr
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.)
Carl Linnaeus, Erasmus Darwin and Anna Seward: Botanical Poetry and Female Education

Science.gov (United States)

George, Sam

2014-03-01

This article will explore the intersection between `literature' and `science' in one key area, the botanical poem with scientific notes. It reveals significant aspects of the way knowledge was gendered in the Enlightenment, which is relevant to the present-day education of girls in science. It aims to illustrate how members of the Lichfield Botanical Society (headed by Erasmus Darwin) became implicated in debates around the education of women in Linnaean botany. The Society's translations from Linnaeus inspired a new genre of women's educational writing, the botanical poem with scientific notes, which emerged at this time. It focuses in particular on a poem by Anna Seward and argues that significant problems regarding the representation of the Linnaean sexual system of botany are found in such works and that women in the culture of botany struggled to give voice to a subject which was judged improper for female education. The story of this unique poem and the surrounding controversies can teach us much about how gender impacted upon women's scientific writing in eighteenth century Britain, and how it shaped the language and terminology of botany in works for female education. In particular, it demonstrates how the sexuality of plants uncovered by Linnaeus is a paradigmatic illustration of how societal forces can simultaneously both constrict and stimulate women's involvement in science. Despite the vast changes to women's access in scientific knowledge of the present day, this `fair sexing' of botany illustrates the struggle that women have undergone to give voice to their botanical knowledge.
The 5th surveillance testing for Kori unit 1 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)

2000-08-01

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 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)
Monitoring of primary circuit and reactor of NPP A-1

International Nuclear Information System (INIS)

Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

2005-01-01

Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)
Characterisation of reactor control rod drives. Specification 1-6. Reaktorstellstabantriebe. Typenblaetter 1-6

Energy Technology Data Exchange (ETDEWEB)

1975-03-01

The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN).

The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

International Nuclear Information System (INIS)

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

2004-01-01

The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)
Comparison of analysis methods for burnup credit applications

International Nuclear Information System (INIS)

Sanders, T.L.; Brady, M.C.; Renier, J.P.; Parks, C.V.

1989-01-01

The current approach used for the development and certification of spent fuel storage and transport casks requires an assumption of fresh fuel isotopics in the criticality safety analysis. However, it has been shown that there is a considerable reactivity reduction when the isotopics representative of the depleted (or burned) fuel are used in a criticality analysis. Thus, by taking credit for the burned state of the fuel (i.e., burnup credit), a cask designer could achieve a significant increase in payload. Accurate prediction of k eff for spent fuel arrays depends both on the criticality safety analysis and the prediction of the spent fuel isotopics via a depletion analysis. Spent fuel isotopics can be obtained from detailed multidimensional reactor analyses, e.g. the code PDQ, or from point reactor burnup models. These reactor calculations will help verify the adequacy of the isotopics and determine Δk eff biases for various analysis assumptions (with and without fission products, actinide absorbers, burnable poison rods, etc.). New software developed to interface PDQ multidimensional isotopics with KENO V.a reactor and cask models is described. Analyses similar to those performed for the reactor cases are carried out with a representative burnup credit cask model using the North Anna fuel. This paper presents the analysis methodology that has been developed for evaluating the physics issues associated with burnup credit. It is applicable in the validation and characterization of fuel isotopics as well as in determining the influence of various analysis assumptions in terms of δk eff . The methodology is used in the calculation of reactor restart criticals and analysis of a typical burnup credit cask
Computer codes for simulation of Angra 1 reactor steam generator

International Nuclear Information System (INIS)

Pinto, A.C.

1978-01-01

A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt
Civilian Power Program. Part 1, Summary, Current status of reactor concepts

Energy Technology Data Exchange (ETDEWEB)

Author, Not Given

1959-09-01

This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).
‘The Good Terrorist(s’? Interrogating Gender and Violence in Ann Devlin’s ‘Naming the Names’ and Anna Burns’ No Bones

Directory of Open Access Journals (Sweden)

Fiona McCann

2012-03-01

Full Text Available This paper aims to analyse the depiction of IRA female volunteers in Ann Devlin’s “Naming the Names” (1986 and Anna Burns’ No Bones (2001 and to consider the relationship established between gender and violence in these texts. I investigate the extent to which the female terrorists portrayed conform to the “mother, monster, whore” paradigm identified by Laura Sjoberg and Caron Gentry (2007 in their study of women’s violence in global politics and consider what differences, if any, are established with these characters’ male counterparts. The ways in which both authors destabilise traditional gender stereotypes is also explored, as is the question of whether these texts might be considered as feminist fictions.
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.
Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

International Nuclear Information System (INIS)

Duque O, J.; Velez A, L.H.

1987-01-01

Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it
Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1973-01-01

By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)
Characterisation of reactor control rod drives. Specification 1-6

International Nuclear Information System (INIS)

1975-03-01

The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de
Developing maintainability in controlled thermonuclear reactors. Progress report, October 1, 1977--April 30, 1978

International Nuclear Information System (INIS)

Zahn, H.S.

1977-05-01

During the period 1 October 1977 through 30 April 1978 the study has completed work on Task 6, Candidate Reference Systems. Four candidate reference systems have been defined. These are based on the conceptual designs of the UWMAK-III, the General Atomic Company Demonstration Power Reactor, the Oak Ridge National Laboratory Cassette defined in the Demonstration Power Study and the Culham laboratory Mark II Reactors. These reactor concepts are normalized to 3000 MW/sub th/ and near minimum cost of electricity. In addition, designs of four major subsystems have been selected and defined for application to these reactors. These include a primary coolant system, primary and secondary vacuum zone systems, the neutral beam injection system and the magnetic field system. These magnet systems are unique to each reactor. The cases for which maintenance plans are being developed in Task 7 have been selected to allow evaluation of design features, particularly the vacuum wall locations, and the impacts of unscheduled and contact maintenance of subsystems on the cost of electricity
The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

1999-01-01

The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)
TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors
Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that
Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

2009-01-01

The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)
Probabilistic risk analysis of Angra-1 reactor

International Nuclear Information System (INIS)

Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

1986-01-01

The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt
Fusion reactor design and technology 1986. V. 1

International Nuclear Information System (INIS)

1987-01-01

The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs
Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

International Nuclear Information System (INIS)

Ashoub, N.; Saleh, H.

1995-01-01

The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)
Motor proficiency profile of Grade 1 learners in the North West ...

African Journals Online (AJOL)

This study aimed to establish a comprehensive profile of the motor proficiency of Grade 1 learners in the North West Province of South Africa, taking into account gender and racial differences and strengths and weaknesses. A stratified randomised sample of 816 Grade 1 learners (419 boys, 397 girls, mean age 6.84 years ...
Integral tightness measurements at the Paks-1 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

1983-01-01

The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.
Molten-salt reactor strategies viewed from fuel conservation effect, (1)

International Nuclear Information System (INIS)

Furuhashi, Akira

1976-01-01

Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

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