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Sample records for north anna-4 reactor

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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, ...

  10. 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

  11. 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.

  12. 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

  13. [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.

  14. 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

  15. 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

  16. 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

  17. 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"

  18. 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.

  19. 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 ...

  20. 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).

  1. 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.

  2. [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.

  3. 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…

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. [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

  9. 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.

  10. 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.

  11. 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).

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. Fuel recycling and 4. generation reactors

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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 patient’s 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.)

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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.)

  14. 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

  15. Intertekstualiteit en die Bose in Kroniek van Perdepoort (Anna M. Louw

    Directory of Open Access Journals (Sweden)

    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.

  16. 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.

  17. 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.

  18. 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

  19. 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...

  20. Submersion-Subcritical Safe Space (S4) reactor

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  1. 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

  2. [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.

  3. 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.

  4. Vos, Faust, Voss: raakpunte tussen Vos deur Anna M. Louw en enkele ander tekste

    Directory of Open Access Journals (Sweden)

    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.

  5. 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.

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

    International Nuclear Information System (INIS)

    Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

    2014-01-01

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

  7. Abjeção em textos de André Sant’Anna

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-01

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

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. “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...

  16. 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)

  17. 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)

  18. 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

  19. 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...

  20. 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

  1. 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.

  2. JENDL-4.0 benchmarking for fission reactor applications

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  3. 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...

  4. Latin and knitting. Educational and biographical experiences in «My life» by Anna Franchi

    Directory of Open Access Journals (Sweden)

    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

  5. 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.

  6. 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

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

    International Nuclear Information System (INIS)

    Tombakoglu, M.; Cecen, Y.

    2001-01-01

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

  8. 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

  9. 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...

  10. 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)

  11. Chernobyl: recovery operations and the entombment of Reactor 4

    International Nuclear Information System (INIS)

    Dalziel, S.P.C.

    1988-01-01

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

  12. 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

  13. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

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

  14. 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

  15. Formation of Barents Sea Branch Water in the north-eastern Barents Sea

    Directory of Open Access Journals (Sweden)

    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.

  16. 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…

  17. 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

  18. 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.

  19. Accounting history and the feminine gender: The case Anna Jansen, Queen of Maranhão (19th Century

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. 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

  3. Nuclear energy. The innovations of the N4 reactor

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

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

  4. 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...

  5. 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…

  6. 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.

  7. Depleted Reactor Analysis With MCNP-4B

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. Methane (CH4) Flux for North America L4 Daily V1 (CMS_CH4_FLX_NAD) at GES DISC

    Data.gov (United States)

    National Aeronautics and Space Administration — The CMS Methane (CH4) Flux for North America data set contains estimates of methane emission in North America based on an inversion of the GEOS-Chem chemical...

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

    International Nuclear Information System (INIS)

    Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

    2010-01-01

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

  15. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  16. 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.

  17. 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…

  18. Dr. Anna G. Jonasdottir: Acceptance Speech for Honorary Doctorate from Faculty of Political Science, University of Iceland. Given 18th of June, 2015

    Directory of Open Access Journals (Sweden)

    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.

  19. ["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.

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

    International Nuclear Information System (INIS)

    Liu Guisheng

    1999-01-01

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

  1. 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.

  2. Review Essay: John Pratt and Anna Eriksson (2013 'Contrasts in Punishment: An Explanation of Anglophone Excess and Nordic Exceptionalism'

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.

    1993-12-01

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

  4. 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

  5. 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

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

    International Nuclear Information System (INIS)

    Platonov, P.A.

    2007-01-01

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

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

    Science.gov (United States)

    1992-06-01

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

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

    International Nuclear Information System (INIS)

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

    1983-06-01

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

  9. THE FEMALE’S MASCULINITIES AGAINST SIAMESE PATRIARCHAL SYSTEMS IN ELIZABETH HAND’S ANNA AND THE KING

    Directory of Open Access Journals (Sweden)

    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.

  10. 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

    Directory of Open Access Journals (Sweden)

    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

  11. 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

  12. Fast breeder reactors--lecture 4

    International Nuclear Information System (INIS)

    Marshall, W.; Davies, L.M.

    1986-01-01

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

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

    Science.gov (United States)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. 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

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

    International Nuclear Information System (INIS)

    1984-06-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  17. 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.



    4pt" class="MsoNormal">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.

  18. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  20. 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.

  1. 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.

  2. 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.

  3. Questions about the reactor accident with Chernobyl-4

    International Nuclear Information System (INIS)

    Heijboer, R.J.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Orso, J; Marenzana, A

    2012-01-01

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

  5. 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.

  6. 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)

  7. 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.)

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

    International Nuclear Information System (INIS)

    1992-07-01

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

  9. [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.

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  17. 9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

    Science.gov (United States)

    2010-01-01

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

  18. Present status of fusion reactor materials, 4

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  19. 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.

  20. 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.

  1. 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.

  2. 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...

  3. 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.

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

    Science.gov (United States)

    Hoogenboom, J. Eduard; Sjenitzer, Bart L.

    2014-06-01

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

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

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    2013-01-01

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

  6. 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

  7. 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)

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

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

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

  9. 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.

  10. 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.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Fuster, S.; Tarabella, A.

    1967-01-01

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

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

    International Nuclear Information System (INIS)

    Cognet, G.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  16. 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

  17. 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.

  18. 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...

  19. An ultracold neutron source at the NC State University PULSTAR reactor

    Science.gov (United States)

    Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

    2007-08-01

    Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

  20. 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?

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

    International Nuclear Information System (INIS)

    2014-02-01

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

  2. 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.

  3. 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?

  4. 15 CFR 746.4 - North Korea.

    Science.gov (United States)

    2010-01-01

    ... entertainment software and equipment; recreational sports equipment; tobacco; wine and other alcoholic beverages... items meeting subsistence needs) intended for the benefit of the North Korean people; items in support...

  5. 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.

  6. Floods of July 4-8, 1969, in north-central Ohio

    Science.gov (United States)

    Mayo, Ronald I.; Webber, Earl E.; Ellis, Davis W.

    1971-01-01

    The storm of July 4-5, 1969, in north-central Ohio was an unprecedented event; never before has such intense and widespread precipitation been recorded for a summer storm in Ohio (U.S. Dept. of Commerce, 1969). More than 14 inches of rainfall in less than 24 hours were observed at several places. In areal extent more than 4 inches of rainfall occurred on about 6,000 square miles. Record-breaking floods were observed at many places in north-central Ohio. Of the 50 sites for which the peak discharge was determined 40 are located on unregulated streams. The peak discharge at five of the 40 sites was four times as large as the discharge of the 50-year flood and the peak discharge for 17 sites was more than twice as large as that of the 50-year flood. Severe losses in terms of lives and property damage were experienced; 41 deaths and more than $66 million in property damage were attributed to the rainstorm, accompanying wind, and resulting floods. This report summarizes peak stages and (or) discharges at 55 sites including five reservoirs, in upper Muskingum River basin, in lower Sandusky River basin, and in the Huron River, Vermilion River, and Black River basins.

  7. 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

  8. Safety of research reactors. Topical issues paper no. 4

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  9. Biblioteca escolar das escolas reunidas Sant’anna do Paranahyba/MT (1936-1945: contribuições para o estudo de sua história.

    Directory of Open Access Journals (Sweden)

    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

  10. 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

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

    International Nuclear Information System (INIS)

    Campioni, Guillaume; Mounier, Claude

    2006-01-01

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

  12. 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

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

    International Nuclear Information System (INIS)

    2014-02-01

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

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

    International Nuclear Information System (INIS)

    2007-03-01

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

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

    International Nuclear Information System (INIS)

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    International Nuclear Information System (INIS)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  20. 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.

  1. 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.

  2. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

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

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

    International Nuclear Information System (INIS)

    1967-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  6. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    2009-08-01

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

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

    International Nuclear Information System (INIS)

    1996-09-01

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

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

    International Nuclear Information System (INIS)

    Foulquier, H.; Clara, P.

    1964-01-01

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

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

    International Nuclear Information System (INIS)

    Behar, Christophe

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2012-03-01

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

  11. Canadian supercritical water reactor modeling using G4STORK

    International Nuclear Information System (INIS)

    Ford, W.; Buijs, A.

    2015-01-01

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

  12. Canadian supercritical water reactor modeling using G4STORK

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2012-01-01

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

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

    Science.gov (United States)

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

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

  17. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    1984-11-01

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

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

    International Nuclear Information System (INIS)

    1984-04-01

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

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

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

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

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

    International Nuclear Information System (INIS)

    Motte, F.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

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

  4. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

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

  6. CMS (Carbon Monitoring System) Methane (CH4) Flux for North America 0.5 degree x 0.667 degree V1 (CMS_CH4_FLX_NA) at GES DISC

    Data.gov (United States)

    National Aeronautics and Space Administration — The CMS Methane (CH4) Flux for North America data set contains estimates of methane emission in North America based on an inversion of the GEOS-Chem chemical...

  7. 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.

  8. 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.

  9. 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

  10. Nuclear reactor technology progress report, vol. 4

    International Nuclear Information System (INIS)

    1981-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

    Guentay, S.

    1996-12-01

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

  13. The accident of Chernobylsk-4 reactor and its consequences

    International Nuclear Information System (INIS)

    1986-01-01

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

  14. 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.)

  15. 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

  16. Growth Comparison of Red Jabon in North Bolaang Mongondow and North Minahasa

    Directory of Open Access Journals (Sweden)

    arif irawan

    2016-04-01

    Full Text Available Red Jabon is priority species that projected to be developed in plantation forests of North Minahasa Regency. Red Jabon is can adapt well various environmental conditions. The natural habitat of the red Jabon in North Sulawesi region was known in North Bolaang Mongondow. This study aimed to compare the growth of red jabon at the age of 1 year in North Minahasa Regency and North Bolaang Mongondow. The experiment was conducted in two (2 locations i.e. in Talawaan Village (North Minahasa Regency and Nunuka Village (North Bolaang Mongondow. Two independent samples t test was conducted to determine the differences in height and diameter growth parameters at two test sites.The results showed that there was no difference in the growth of red Jabon between North Minahasa and North Bolaang Mongondow. The average height and diameter of red Jabon in North Minahasa Regency is at 215.84 cm and 4.34 cm, while the same parameters of North Bolaang Mongondow amounted to 231.79 cm and 4.14 cm. The development efforts on the plantations of red Jabon North Minahasa Regency need to be improved because of the environmental conditions in this region is very suitable for the growth of red Jabon.

  17. 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

  18. 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

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

    Science.gov (United States)

    Kadyk, Thomas; Eikerling, Michael

    2015-08-14

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

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

    International Nuclear Information System (INIS)

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

    2009-06-01

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

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

    International Nuclear Information System (INIS)

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

    2012-03-01

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

  2. 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.

  3. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

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

  5. 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

  6. Decadal changes in CH4 and CO2 emissions on the Alaskan North Slope

    Science.gov (United States)

    Sweeney, C.; Commane, R.; Wofsy, S.; Dlugokencky, E. J.; Karion, A.; Stone, R. S.; Chang, R.; Tans, P. P.; Wolter, S.

    2016-12-01

    Large changes in surface air temperature, sea ice cover and permafrost in the Arctic Boreal Ecosystems (ABE) are significantly impacting the critical ecosystem services and human societies that are dependent on the ABE. In order to predict the outcome of continued change in the climate system of the ABE, it is necessary to look at how past changes in climate have affected the ABE. We look at 30 years of CH4 and 42 years of CO2 observations from the NOAA Global Greenhouse Gas Reference Network site in Barrow, Alaska. By eliminating background trends and only looking at data collected when winds are blowing off the North Slope we find very little change in CH4 enhancements, but significant changes in the CO2 enhancements coming off the tundra. The bulk of both CO2 and CH4 emissions appear to be emitted well after the first snow fall on the North Slope. CO2 emissions are a strongly correlation with summer surface temperatures, while CH4 emissions appear insensitive to the large temperature changes that occurred over the measurement period. These results suggest that CO2, and not CH4 emissions, are a likely pathway for the degradation of permafrost carbon.

  7. 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.

  8. 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.

  9. 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

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

    International Nuclear Information System (INIS)

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

    2013-03-01

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

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

    International Nuclear Information System (INIS)

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

    1986-07-01

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

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

    International Nuclear Information System (INIS)

    1980-01-01

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

  13. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

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

  14. 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

  15. A study on the international cooperation in the nuclear liability system related to the supply of nuclear power plants to North Korea

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Chul Hoon.; Kim, Tae Myeong [The Catholic University, Seoul (Korea)

    2001-12-01

    System of nuclear damage compensation was prepares to protect the interested parties in the implementation of nuclearenterprise and conciliate the conflicts of interests of them. The Light-water Reactor (LWR) Project to supply two units of light-water reactors to North Korea faced difficulties concerning nuclear damage compensation system due to decline of the international reliability and aggravation of economic condition of North Korea. It is necessary to study the special nuclear damage compensation system of the LWR Project to promote the Project and peaceful uses of atomic energy in northeast Asia. The contents and scope of the study is composed as follows; 1. Background of the LWR Project . the pending issues of them and the necessary of the special nuclear damage compensation system 2. Investigation of nuclear damage compensation system of United States, Japan, German, France and Korea 3. Account of conventions on liability for nuclear damage, especially Vienna Convention and its Protocols 4. Searching for issues of the nuclear damage compensation system of the LWR Project and its resolution 5. Comprehensive arrangement on the main issues through the study. 4 tabs. (Author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Subagjo

    2014-07-01

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

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

    International Nuclear Information System (INIS)

    Karriem, Z.; Zamonsky, O.M.

    2014-01-01

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

  19. 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

  20. 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.

  1. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  3. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    Dupraz, B.; Bertel, E.

    2003-01-01

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

  4. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-15

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

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

    Science.gov (United States)

    Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

    2011-03-15

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

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

    International Nuclear Information System (INIS)

    Van Heek, A.I.

    1996-06-01

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

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  11. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  12. 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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. 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

  15. 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.

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

    International Nuclear Information System (INIS)

    Arndt, D.; Schmidt, W.

    1989-01-01

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

  17. 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

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

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

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

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

    International Nuclear Information System (INIS)

    Setiyanto; Sembiring, Tagor M.; Pinem, Surian

    2007-01-01

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

  20. Elements on reactor control

    International Nuclear Information System (INIS)

    Bruna, G.B.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

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

  2. Supercritical Water Reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  3. 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)

  4. High risk association of IL-4 VNTR polymorphism with asthma in a North Indian population.

    Science.gov (United States)

    Birbian, Niti; Singh, Jagtar; Jindal, Surinder Kumar; Sobti, Ranbir Chander

    2014-03-01

    A case-control study was conducted to evaluate the role of IL-4 VNTR polymorphism in asthma that has been associated with various inflammatory diseases worldwide. This is the first case-control study conducted in India, investigating the role of IL-4 VNTR polymorphism in asthma pathogenesis. A case-control study was performed with a total of 824 adult subjects, inducting 410 asthma patients and 414 healthy controls from North India. The genotypes were identified by polymerase chain reaction. Statistical analysis for the IL-4 VNTR polymorphism revealed that the Rp1 allele was significantly associated with asthma with OR=1.47, 95% CI (1.11-1.94) and p=0.005. The Rp1/Rp1 homozygous mutant genotype posed a high risk towards asthma with OR=2.39, 95% CI (0.96-6.14) and p=0.040. The Rp2/Rp1 heterozygous genotype also posed a risk towards asthma with OR=1.39, 95% CI (1.00-1.94) and p=0.040. Most of the phenotypic traits were significantly associated with the disease. IL-4 VNTR polymorphism is a high risk factor for asthma in the studied North Indian population. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    2012-12-01

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

  6. 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.

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

    International Nuclear Information System (INIS)

    Yang Fanglin; Yang Yingli; Wang Guangshen

    1989-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-15

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

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

    International Nuclear Information System (INIS)

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

    2014-03-01

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

  10. The molten salt reactor adventure

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1985-01-01

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

  11. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

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

  12. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  13. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

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

    1974-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. 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)

  17. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

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

  18. Compact torsatron reactors

    International Nuclear Information System (INIS)

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

    1988-05-01

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

  19. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  1. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rumis, D; Leszczynski, F

    1991-12-31

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

  3. 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.

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

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

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

  5. The IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

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

  6. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

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

  7. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

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

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

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

  9. 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.

  10. 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

  11. Studies of conceptual spheromak fusion reactors

    International Nuclear Information System (INIS)

    Katsurai, M.; Yamada, M.

    1982-01-01

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

  12. 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

  13. ‘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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  15. Directory of Nuclear Research Reactors 1994

    International Nuclear Information System (INIS)

    1995-08-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  17. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-15

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

  18. 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

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-12-15

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

  2. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    1995-02-01

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

  3. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  5. Molten salt reactors and the oil sands: odd couple or key to north american energy independence?

    Energy Technology Data Exchange (ETDEWEB)

    LeBlanc, D., E-mail: d_leblanc@rogers.com [Ottawa Valley Research Associates Ltd., Ottawa, Ontario (Canada); Quesada, M.; Popoff, C.; Way, D. [Penumbra Energy, Calgary, Alberta (Canada)

    2012-07-01

    The use of nuclear power to aid oil sands development has often been proposed largely due to the virtual elimination of natural gas use and thus a large reduction in GHG emissions. Nuclear power can replace natural gas for process steam production (SAGD) and electricity generation but also potentially for hydrogen production to upgrade bitumen for pipeline transit, synthetic crude production and even at the final refinery stage. Prior candidates included CANDU and gas cooled Pebble Bed Reactors. The case for CANDU use can be shown to be marginally economic with a proven technology but with an uncertainty of current construction costs and too large a unit size (~2400 MWth). PBRs offered modest theoretical cost savings, smaller unit size and the ability to offer higher temperatures needed for thermochemical hydrogen production from water. Interest in PBRs however has greatly waned with the cancellation of their major South African development program which highlighted the severe challenges of helium as a coolant and TRISO fuel manufacturing. More recently, Small Modular Reactors based on scaled down light water reactor technology have attracted interest but are unlikely to compete economically outside of niche applications. However, a 'new' reactor option, the Molten Salt Reactor, has been rapidly gaining momentum over the past decade. This 'new' technology was actually developed over 50 years ago as a thorium breeder reactor to compete with the sodium cooled fast breeder reactor (U-Pu cycle). During this time two molten salt test reactors were constructed. A modern version however would likely be a simpler converter design using Low Enriched Uranium but needing only a small fraction the uranium resources of LWRs or CANDUs. Besides resource sustainability, these unique designs offer large potential improvements in the areas of capital costs, safety and nuclear waste. This presentation will explain the unique attributes and advantages of these

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  7. MMRW-BOOKS, Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors

    International Nuclear Information System (INIS)

    Williams, M.M.R.

    2007-01-01

    Description: Prof. M.M..R Williams has now released three of his legacy books for free distribution: 1 - M.M.R. Williams: The Slowing Down and Thermalization of Neutrons, North-Holland Publishing Company - Amsterdam, 582 pages, 1966. Content: Part I - The Thermal Energy Region: 1. Introduction and Historical Review, 2. The Scattering Kernel, 3. Neutron Thermalization in an Infinite Homogeneous Medium, 4. Neutron Thermalization in Finite Media, 5. The Spatial Dependence of the Energy Spectrum, 6. Reactor Cell Calculations, 7. Synthetic Scattering Kernels. Part II - The Slowing Down Region: 8. Scattering Kernels in the Slowing Down Region, 9. Neutron Slowing Down in an Infinite Homogeneous Medium, 10.Neutron Slowing Down and Diffusion. 2 - M.M.R. Williams: Mathematical Methods in Particle Transport Theory, Butterworths, London, 430 pages, 1971. Content: 1 The General Problem of Particle Transport, 2 The Boltzmann Equation for Gas Atoms and Neutrons, 3 Boundary Conditions, 4 Scattering Kernels, 5 Some Basic Problems in Neutron Transport and Rarefied Gas Dynamics, 6 The Integral Form of the Transport Equation in Plane, Spherical and Cylindrical Geometries, 7 Exact Solutions of Model Problems, 8 Eigenvalue Problems in Transport Theory, 9 Collision Probability Methods, 10 Variational Methods, 11 Polynomial Approximations. 3 - M.M.R. Williams: Random Processes in Nuclear Reactors, Pergamon Press Oxford New York Toronto Sydney, 243 pages, 1974. Content: 1. Historical Survey and General Discussion, 2. Introductory Mathematical Treatment, 3. Applications of the General Theory, 4. Practical Applications of the Probability Distribution, 5. The Langevin Technique, 6. Point Model Power Reactor Noise, 7. The Spatial Variation of Reactor Noise, 8. Random Phenomena in Heterogeneous Reactor Systems, 9. Associated Fluctuation Problems, Appendix: Noise Equivalent Sources. Note to the user: Prof. M.M.R Williams owns the copyright of these books and he authorises the OECD/NEA Data Bank

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

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1976-09-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Directory of Open Access Journals (Sweden)

    S. N. Rokhman

    2015-04-01

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

  12. Spatial kinetics in nuclear reactor systems. Chapter 4

    International Nuclear Information System (INIS)

    Owens, D.H.

    1980-01-01

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

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

    Science.gov (United States)

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

    2013-06-14

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

  14. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-09-01

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

  16. Proceedings of the 4th international symposium on material testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Suzuki, Masahide

    2012-03-01

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

  17. Proceedings of the 4th international symposium on material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

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

    International Nuclear Information System (INIS)

    Ardani

    2011-01-01

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

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

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-30

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-15

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

  3. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  4. Cermet fuel reactors

    International Nuclear Information System (INIS)

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

    1987-09-01

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

  5. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Jammes, Ch.

    2010-07-01

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

  7. North Korea's nuclear weapons development. Implications for future policy

    International Nuclear Information System (INIS)

    Pollack, J.D.

    2010-01-01

    This essay assesses North Korea's long-standing quest for nuclear weapons; alternative strategies for inhibiting Pyongyang's weapons development; and the potential implications for regional security and nonproliferation should the Democratic People's Republic of Korea (DPRK) retain and enhance its weapons programs. North Korea's pursuit of a nuclear weapons capability has long provoked heated debate among policy makers and research analysts about the purposes of engagement with the North, reflecting the repeated frustrations in efforts to negotiate Korean denuclearization. These debates reflect widely divergent views of the North Korean regime; its sustainability as an autonomous political, economic, and military system; and the potential consequences of continued nuclear development in this isolated, highly idiosyncratic state. These questions assume additional salience as North Korea approaches a leadership succession for only the second time in its six-decade history. The effort to inhibit North Korea's pursuit of nuclear weapons is among the longest running and least successful sagas in international security and non-proliferation policy of the past quarter century. In early 2010, Pyongyang claims a rudimentary nuclear capability by possession of weaponized plutonium, the conduct of two nuclear tests, and advances in the production of enriched uranium as an alternative means of fissile material production, though the latter step is nominally justified as a source for reactor fuel. North Korea defends its pursuit of a nuclear deterrent to counter what Pyongyang deems existential threats posed by the United States.Despite the resumption of high-level diplomatic contact between Washington and Pyongyang in late 2009, realization of a non-nuclear Korean Peninsula remains a very remote prospect. The DPRK insists that a peace agreement between the U.S. and North Korea and hence the cessation of 'hostile DPRK-U.S. relations' are necessary before any consideration of

  8. The integral fast reactor concept

    International Nuclear Information System (INIS)

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

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  9. Reactor safety in Eastern Europe. Proceedings

    International Nuclear Information System (INIS)

    1995-02-01

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

  10. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

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

  11. The North-Korean-Syrian Connection: Take It Seriously But Address It Wisely

    International Nuclear Information System (INIS)

    Santoro, David

    2008-01-01

    The release of photographs of what U.S. officials say was a Syrian nuclear reactor built with North Korean assistance (and destroyed by Israel in September 2007) could undermine the success of ongoing negotiations with Pyongyang about its nuclear weapons activities. In addition to pressing Syria to admit its nuclear ambitions, the Bush Administration argues that the disclosure of these photographs will pressure Pyongyang to honor its lagging non-proliferation commitments laid out in a six-party agreement reached in February 2007. The problem is that this could also bring current negotiations to a breakdown. In other words, in the face of U.S. confrontation, North Korean officials could simply decide to walk out of the negotiating table altogether. In the recent past, the absence of a negotiating process with Pyongyang had dramatic consequences. Following the collapse of negotiations in 2002 (when Washington confronted Pyongyang with an alleged clandestine uranium enrichment program), North Korea moved swiftly to unfreeze its plutonium production facilities, expel international inspectors, withdraw from the Nuclear Non-Proliferation Treaty (NPT), restart its reprocessing activities and, finally, conduct a nuclear test in October 2006. During this period, the possibility of war and regional proliferation increased and many proliferation experts worried that the financially strapped Communist regime would attempt to sell its nuclear technology to anyone willing to buy it, including terrorist groups. A negotiating process was finally restored with the February 2007 agreement, which the Bush Administration hailed as a major breakthrough. Nearly fifteen months down the track, Pyongyang has closed and almost entirely disabled its main nuclear reactor (Yongbyon). But it has missed key deadlines to provide detailed information on its stocks of plutonium and nuclear weapons, its uranium enrichment program as well as a complete accounting of its past proliferation activities

  12. Fusion reactor materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

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

  13. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-15

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

  14. Mirror hybrid reactor optimization studies

    International Nuclear Information System (INIS)

    Bender, D.J.

    1976-01-01

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

  15. 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 ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  17. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  18. The innovation of the subspecialty of Paediatric Virology: An interview with Research Professor of Molecular Virology Anna Kramvis.

    Science.gov (United States)

    Mammas, Ioannis N; Spandidos, Demetrios A

    2017-10-01

    Professor Anna Kramvis, Research Professor of Molecular Virology at the University of the Witwatersrand in Johannesburg, South Africa, talks about direct-acting antiviral treatments against hepatitis C virus (HCV), as well as the perspective of the development of an effective vaccine against HCV. She emphasises the necessity of vaccination against hepatitis B virus (HBV), highlighting that it is very important that vaccination should be administered at birth in order to prevent mother-to-child transmission (MTCT) of HBV. Professor Kramvis states that vaccination against HBV is safe and that HBV and HCV infections are not contraindications for breastfeeding. Regarding the challenge of Paediatric Virology, she believes that it is a field that during the last years is increasing exponentially, while she concurs that Paediatric Virology subspecialty will be a popular choice for infectious diseases subspecialists. In the context of the 3rd Workshop on Paediatric Virology, which will be held in Athens on October 7th, 2017, Professor Kramvis will give her key lecture on MTCT of HBV and HCV.

  19. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

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

  20. Molten salt reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. Relación entre la variación de las condiciones ambientales y la temperatura interna de la yema del manzano 'Anna' Inner bud temperature of' Anna' apple trees in relation to varying environmental conditions

    Directory of Open Access Journals (Sweden)

    Casierra Posada Fannor

    2001-12-01

    Full Text Available En Paipa (Colombia, se realizó un ensayo de campo con el fin de determinar el efecto de la variación de las condiciones ambientales sobre la temperatura interna de la yema del manzano 'Anna' cultivado en condiciones de altiplano tropical. Los registros se hicieron en un árbol de cuatro años, el cual se defolió manualmente y se podó pocas semanas después de la cosecha. Se midió la temperatura del aire y se comparó con la temperatura de una yema en el árbol. Para medir la temperatura de la yema se colocó un sensor digital muy delgado entre los catáfilos de la yema; la temperatura ambiente se registró con un termómetro de bulbo. Las mediciones se realizaron cada media hora entre las 7:00 y las 17:30 horas del día, durante 19 días, empezando el día en que el árbol fue defoliado hasta cuando la yema alcanzó el estado fenológico "green tip". Al tiempo con cada medición de temperatura se registró el fenómeno climático más sobresaliente (lluvia, tiempo soleado, viento, tiempo nublado, etc. y los datos se agruparon según cada fenómeno climático. Los resultados mostraron que la yema presentó una temperatura 1,49°C por encima de la temperatura del aire durante el transcurso del día, en la noche no hubo diferencia de temperatura. En condiciones de lluvia, la temperatura
    de la yema estuvo 0,57°C por encima de la temperatura
    del aire. Bajo cielo nublado la yema estuvo 0,56°C más cálida
    que el aire. En condiciones de viento la yema presentó una
    diferencia de 0,87°C sobre la temperatura del aire. En días
    soleados, la yema mantuvo una temperatura 2,3°C sobre la temperatura del aire. Las diferencias entre la temperatura del aire y la de la yema se discuten con base en la transpiración, la convección y la radiación.A field experiment was conducted in Paipa (Colombia to study the effect of environmental varying over inner bud temperature of an 'Anna' apple tree 4 years old growing in tropical

  2. Possibilities of TWR and long life reactor

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Shimazu, Yoichiro; Handa, Norihiko

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    1984-06-01

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

  4. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  5. 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

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

    International Nuclear Information System (INIS)

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

    2017-02-01

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

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

    International Nuclear Information System (INIS)

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

    2015-02-01

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

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

    NARCIS (Netherlands)

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

    2010-01-01

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

  9. 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

    Directory of Open Access Journals (Sweden)

    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.

  10. Kinetic analysis of sub-prompt-critical reactor assemblies

    International Nuclear Information System (INIS)

    Das, S.

    1992-01-01

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

  11. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

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

  12. Linguagem experimental: o discurso do consumo em Sexo, de André Sant'Anna, e o discurso da 'existencialidade' , em Lugar público, de José Agrippino de Paula

    OpenAIRE

    Romulo de Almeida Portella

    2006-01-01

    O objetivo principal desta dissertação é investigar as obras Sexo, de André SantAnna, e Lugar Público, de José Agrippino de Paula, a partir das noções que lidam com a vitalidade das formulações verbais, no que dizem respeito à elaboração da linguagem literária e a estruturação desta como conhecimento a ser apreendido. O trabalho contribui para um melhor entendimento da estética do experimentalismo, que é abordada em tais narrativas ficcionais. This dissertations main aim is to inves...

  13. Mirror reactor studies

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Pierre, J.R.M.

    1996-01-01

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

  15. Turning points in reactor design

    International Nuclear Information System (INIS)

    Beckjord, E.S.

    1995-01-01

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

  16. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1975-07-01

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

  18. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Radionuclide concentration in ground-level air from 1986 to 1987 in North Germany and North Norway

    International Nuclear Information System (INIS)

    Kolb, W.

    1988-03-01

    The activity concentration of various fission products and some other radionuclides (e.g. Be-7, Na-22, K-40 and Pb-210) contained in surface air were determined by γ-spectroscopy. The mean monthly acitvity concentrations of up to 30 radionuclides measured in 1986 and 1987 in Brunswick, Berlin and Skibotn (North Norway) are tabulated. The Chernobyl accident of April 26, 1986, resulted in 1986 in an annual mean Cs-137 activity concentration of 2.4 mBq/m 3 in Brunswick, 8.8 mBq/m 3 in Berlin and 0.3 mBq/m 3 at Skibotn. In 1987 the Cs-137 concentrations were just about 1% of these values. Occasionally fresh fission products from other sources were detected as e.g. I-131 in March 1987 (very likely released from a reactor site in Ukraine) and in August 1987 (released from an underground nuclear test on Novaja Zemlya together with other short-lived fission products). The effective dose equivalent due to inhalation of fission products is estimated for all three sites and compared with the Pb-210 inhalation dose. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  2. Prospect of realizing nuclear fusion reactors

    International Nuclear Information System (INIS)

    1989-01-01

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

  3. 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

  4. Reactor Physics Behind the Chernobyl Accident

    International Nuclear Information System (INIS)

    Reisch, F.

    1999-01-01

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

  5. 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

  6. Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

    1990-05-01

    Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  8. Distributional changes of American martens and fishers in eastern North America, 1699-2001: Chapter 4

    Science.gov (United States)

    Krohn, William B.

    2012-01-01

    Contractions in the geographic distributions of the American marten ( Martes americana) and fi sher ( M. pennanti) in eastern North America south of the St. Lawrence River between Colonial times (ca. 1650–1800) and the fi sher’s recent range expansion (ca. 1930–present) are well documented, but causal factors in these range contractions have only partially been studied. Traditional explanations for range contractions by both species are forest clearing and unregulated trapping; little consideration has been given to alternative explanations. It has been hypothesized that deep snow limits the distribution of fi shers, and that high fi sher populations limit the distribution of martens. I assessed the potential contributions of these factors to observed range contractions for these species by evaluating expected patterns of change in their historical distributions since Colonial times. Using published data on the distribution of martens and fi shers in eastern North America, including early and contemporary fur-harvest records ( n = 60,702), I found that broad-scale changes in their geographic distributions in eastern North America were consistent with 3 of those expectations, and partially so with a 4th. I recognize that retrospective analyses cannot establish the relative importance of land clearing, unregulated trapping, and changing climatic conditions on observed range contractions; nevertheless, when historical data from eastern North America are viewed in the context of long-term climate warming and the results of recent ecological studies, they suggest that traditional arguments may only partially explain historical range contractions for both species. This study further suggests that under a warming climate, northern range boundaries for the fi sher will expand, and southern range boundaries for the American marten will continue to contract.

  9. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    DEFF Research Database (Denmark)

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

    2013-01-01

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

  11. Infantile anorexia, co-excitation and co-mastery in the parent/baby cathexis: The contribution of Sigmund and Anna Freud.

    Science.gov (United States)

    Cascales, Thomas

    2017-04-01

    Recent epidemiological studies show that 2% of babies in ordinary paediatric clinics suffer from infantile anorexia. In the first part of this paper we present a case study from our hospital clinical activity. Our framework combines clinical psychoanalytic sessions and perinatal videos. In the second part, we will focus on the concepts of instinct and excitation proposed by Sigmund Freud and the concept of mastery proposed by Anna Freud. In the third part, we will examine these concepts in the light of the case study. The fourth part is devoted to clinical recommendations from our hospital psychoanalytic practice. In conclusion, unlike other clinical settings, the psychoanalytic setting allows for the elaboration of the parental hatred included in the libidinal cathexis. Our psychoanalytic setting (sessions/videos) makes it possible to decontaminate parental intrapsychic elements overloaded with excitement, saturated with hate elements, and rendered sterile by the instinct for mastery. An initial part of the treatment process involves working through the intersubjective elements observed in the video. Copyright © 2016 Institute of Psychoanalysis.

  12. Cryogenic system design for a compact tokamak reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    1973-01-01

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

  14. North Korea's nuclear weapons program:verification priorities and new challenges.

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Duk-ho (Korean Consulate General in New York)

    2003-12-01

    A comprehensive settlement of the North Korean nuclear issue may involve military, economic, political, and diplomatic components, many of which will require verification to ensure reciprocal implementation. This paper sets out potential verification methodologies that might address a wide range of objectives. The inspection requirements set by the International Atomic Energy Agency form the foundation, first as defined at the time of the Agreed Framework in 1994, and now as modified by the events since revelation of the North Korean uranium enrichment program in October 2002. In addition, refreezing the reprocessing facility and 5 MWe reactor, taking possession of possible weapons components and destroying weaponization capabilities add many new verification tasks. The paper also considers several measures for the short-term freezing of the North's nuclear weapon program during the process of negotiations, should that process be protracted. New inspection technologies and monitoring tools are applicable to North Korean facilities and may offer improved approaches over those envisioned just a few years ago. These are noted, and potential bilateral and regional verification regimes are examined.

  15. 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.)

  16. Astrid (fast breeder nuclear reactor)

    International Nuclear Information System (INIS)

    2014-01-01

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

  17. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

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

  18. Improvement of research reactor sustainability

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. Spatial and temporal patterns of CH4 and N2O fluxes in terrestrial ecosystems of North America during 1979–2008: application of a global biogeochemistry model

    Directory of Open Access Journals (Sweden)

    C. Lu

    2010-09-01

    Full Text Available Continental-scale estimations of terrestrial methane (CH4 and nitrous oxide (N2O fluxes over a long time period are crucial to accurately assess the global balance of greenhouse gases and enhance our understanding and prediction of global climate change and terrestrial ecosystem feedbacks. Using a process-based global biogeochemical model, the Dynamic Land Ecosystem Model (DLEM, we quantified simultaneously CH4 and N2O fluxes in North America's terrestrial ecosystems from 1979 to 2008. During the past 30 years, approximately 14.69 ± 1.64 T g C a−1 (1 T g = 1012 g of CH4, and 1.94 ± 0.1 T g N a−1 of N2O were released from terrestrial ecosystems in North America. At the country level, both the US and Canada acted as CH4 sources to the atmosphere, but Mexico mainly oxidized and consumed CH4 from the atmosphere. Wetlands in North America contributed predominantly to the regional CH4 source, while all other ecosystems acted as sinks for atmospheric CH4, of which forests accounted for 36.8%. Regarding N2O emission in North America, the US, Canada, and Mexico contributed 56.19%, 18.23%, and 25.58%, respectively, to the continental source over the past 30 years. Forests and croplands were the two ecosystems that contributed most to continental N2O emission. The inter-annual variations of CH4 and N2O fluxes in North America were mainly attributed to year-to-year climatic variability. While only annual precipitation was found to have a significant effect on annual CH4 flux, both mean annual temperature and annual precipitation were significantly correlated to annual N2O flux. The regional estimates and spatiotemporal patterns of terrestrial ecosystem CH4 and N2O fluxes in North America generated in this study provide useful information for global change research and policy making.

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

    International Nuclear Information System (INIS)

    Krpelanova, M.; Carny, P.

    2011-01-01

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

  1. Safety philosophy and safety technology of the Soviet RBMK reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

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

  3. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

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

  6. Synchronising EDML and NorthGRIP ice cores using d18O of atmospheric oxygen (d18 Oatm) and CH4 measurements over MIS5 (80-123 kyr)

    DEFF Research Database (Denmark)

    Capron, E; Landais, A; Lemieux-Dudon, B

    2010-01-01

    Water isotope records from the EPICA Dronning Maud Land (EDML) and the NorthGRIP ice cores have revealed a one to one coupling between Antarctic Isotope Maxima (AIM) and Greenland Dansgaard-Oeschger (DO) events back to 50 kyr. In order to explore if this north-south coupling is persistent over...... Marine Isotopic Stage 5 (MIS 5). a common timescale must first be constructed. Here, we present new records of delta O-18 of O-2 (delta O-18(atm)) and methane (CH4) measured in the air trapped in ice from the EDML (68-147 kyr) and NorthGRIP (70-123 kyr) ice cores. We demonstrate that, through the period...... of interest, CH4 records alone are not sufficient to construct a common gas timescale between the two cores. Millennial-scale variations of delta O-18(atm) are evidenced over MIS 5 both on the Antarctic and Greenland ice cores and are coupled to CH4 profiles to synchronise the NorthGRIP and EDML records...

  7. Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Peron, Arthur

    2014-01-01

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

  10. 77 FR 46119 - Notice of Inventory Completion: U.S. Department of the Interior, Bureau of Indian Affairs...

    Science.gov (United States)

    2012-08-02

    ... be culturally affiliated with the human remains should contact Anna Pardo, Museum Program Manager... culturally affiliated with the human remains may contact the U.S. Department of the Interior, Bureau of... Affairs at the address below by September 4, 2012. ADDRESSES: Anna Pardo, Museum Program Manager/NAGPRA...

  11. Corrosion of reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-01-15

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

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

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

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

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

    Science.gov (United States)

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

    2017-09-04

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  16. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  17. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  18. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-07-01

    A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

  19. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-09-01

    A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

  20. Bohemia as the Homeland of the Soul in the Letters of Marina Tsvetaeva to Anna Teskova

    Directory of Open Access Journals (Sweden)

    Massimo Tria

    2014-03-01

    Full Text Available The article examines the letters sent by Marina Tsvetaeva to the translator and public figure Anna Teskova, who was one of the few poet’s Czech intimate friends, whom she met during her stay in Czechoslovakia (1922-1925. In the first part of the paper the Author focuses on the echoes of Prague and its culture in this specific correspondence, trying to ascertain to which extent Marina Tsvetaeva was acquainted with the cultural world she lived in: the lack of knowledge of the Czech language and the limited relationships with Prague intellectuals didn’t allow her to get a deep insight of the Czech cultural milieu. The textual analysis of the letters, though, on the basis of key concepts as “родной”, “родина”, “честь”, confirms the presence of a persistent feeling of Prague-homesickness in Tsvetaeva’s psycho- logical world, which brought her to develop an alternative model of Homeland, substitute to her native Russia and to Germany, which disappointed her after the rise of the Nazi Regime. Through this unique correspondence (including the first part of her Verses dedicated to Bohemia we gain the certainty that Tsvetaeva discovered in Prague and in the Czech lands an idealized model of a pacific, beautiful and faithful Homeland.

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  2. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Science.gov (United States)

    Guidez, Joel; Saturnin, Anne

    2017-11-01

    During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  3. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  4. Demonstrated operational and inherent safety of the prototype fast reactor (PFR)

    International Nuclear Information System (INIS)

    Smedley, J.A.; Gregory, C.V.; Judd, A.M.

    1983-01-01

    The Prototype Fast Reactor (PFR) is sited at Dounreay, on the north coast of Scotland in the United Kingdom, and has been in operation since 1974. Three aspects of the safety of the reactor are described, including the all-important practical consideration of operational safety, a demonstration of the limited consequences of a sodium/water reaction in a steam generator and the ability of the reactor to protect itself against highly improbable incidents. Attention is drawn to the low radiation levels in the plant and the correspondingly low dose rate to personnel. A feature of PFR operation has been the stable and predictable behaviour of its core together with the high degree of reliability exhibited by the engineered safety system. No failures have occurred within the standard driver charge but two experimental fuel pins suffered cladding failure, which was detected easily by the fission gas and delayed neutron detection systems. In the steam generating units sodium and water are separated by the single steel wall of the steam tubes. Although no under-sodium leak has occurred, an experimental programme is continuing and demonstrates that were any such leak to occur its consequences would be containable and would not result in the release of sodium to the environment or any breach of the reactor containment. The final section describes the inherent safety features of the reactor which enable it to survive a range of very improbable incidents even when the engineered safeguards fail. The features considered are natural circulation, which has been demonstrated by reactor experiment; the reactor's negative power coefficient, which, for example, enables the reactor to survive a complete loss of heat sink; and the durability of the fuel pins, demonstrated by a series of boiling experiments in the Dounreay Fast Reactor (DFR). (author)

  5. Fusion reactor materials

    International Nuclear Information System (INIS)

    Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

    1980-01-01

    Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

  6. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

  7. How many reactor accidents will there be

    International Nuclear Information System (INIS)

    Islam, S.; Lindgren, K.

    1986-01-01

    A method for calculation of the probability of nuclear accidents is described. The method is based on the use of data from reactor operating experience, i.e. there have been two major accidents [Three Mile Island and Chernobyl] during 4,000 reactor-years (cumulative operating experience). The authors argue that this method is better than the present ''technical risk assessment'' method based on the likelihood of failure of a reactor component or safety system, used by designers of nuclear reactor. (U.K.)

  8. Preliminary Research on the Verification Task of North Korea's Plutonium Declaration

    International Nuclear Information System (INIS)

    Kim, Hyun Chul; Park, Il Jin

    2009-01-01

    The denuclearization of North Korea seems challenging. North Korea has recognized itself as a nuclear weapon state by carrying out two nuclear tests while many other nations including South Korea have opposed North Korea's nuclear proliferation. As a result of longstanding negotiations, North Korea provided nearly 19,000 pages of operation history of three Yongbyon nuclear facilities on May 8, 2008 and a 60-page declaration of its nuclear activities and programs on June 26, 2008. However, one should notice that declaration documents are by themselves meaningless without their verification. To completely dismantle North Korea's nuclear programs, the verification task based on its declaration documents should be performed very thoroughly, considering the possibility of the presence of the undeclared nuclear materials and facilities. The verification task of North Korea's nuclear declaration consists of many broad themes to deal with, such as the review of declaration documents, the interview with facility operators, the sampling in the field, the laboratory analysis of the sample, data interpretation, and so on. One of the important themes is to verify North Korea's declared plutonium stockpile by comparing the declaration documents with measurement data which can be obtained from the sampling in the field and laboratory analysis. To prepare for the possible future verification of the declared plutonium stockpile, it is meaningful to give a thought on what data can be compared and what samples need to be taken and analyzed. In this study, we focus on the data to be compared and samples to be taken and analyzed for the plutonium accounting, as a preliminary research. To give a quantitative example, the nuclear material of the most recent North Korea's spent fuel rods discharged from the 5 MWe reactor is analyzed. On June 13, 2009, North Korea declared that more than one-third of the spent fuel rods had been reprocessed

  9. Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    The reduction of air pollution has become an international concern over the last ten years because of increases in emissions from mobile and stationary sources. Among these sources, volatile organic compounds (VOC) represent a serious environmental problem, together with NOx, SOx and particulate...... VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

  10. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  11. Safety Evaluation Report related to the renewal of the operating license for the training and research reactor at the University of Michigan (Docket No. 50-2)

    International Nuclear Information System (INIS)

    1985-07-01

    This Safety Evaluation Report for the application filed by the University of Michigan (UM) for renewal of the Ford Nuclear Reactor (FNR) operating license number R-28 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the North Campus of the University of Michigan in Ann Arbor, Michigan. The staff concludes that the reactor can continue to be operated by the University of Michigan without endangering the health and safety of the public

  12. Deposition of RuO 4 on various surfaces in a nuclear reactor containment

    Science.gov (United States)

    Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

    2009-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

  13. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  14. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

  15. Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

    2016-09-26

    In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

  16. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Directory of Open Access Journals (Sweden)

    Guidez Joel

    2017-01-01

    In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  17. Inter renewal travelling wave reactor with rotary fuel columns

    International Nuclear Information System (INIS)

    Terai, Yuzo

    2016-01-01

    To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

  18. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  19. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  2. The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Swam, L.F.P.; Shann, S.H.

    1991-01-01

    This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

  3. Isothermal calorimeter for reactor radiation dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Radak, B; Markovic, V [Institute of Nuclear Sciences Boris Kidric, Odeljenje za radijacionu hemiju, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    An isothermal calorimeter with thermistors for measuring absorbed dose rates from 10{sup 4}-5-6.10{sup 5} rad/h in reactor experimental holes has been designed. A kinetics method for determining the equilibrium temperature difference has been developed, and its application in isothermal calorimetry proved. The expected accuracy in measurements within {+-} 2-5% has been proved by measurements carried out in the reactor. Some data obtained by measurements in the reactor RA are presented (author)

  4. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  5. Compact stellarators as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  6. Proceedings of the advisory committee on reactor safeguards workshop on future reactors

    International Nuclear Information System (INIS)

    2001-12-01

    This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda

  7. Proceedings of the advisory committee on reactor safeguards workshop on future reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-12-01

    This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda.

  8. Rapid preparation of high electrochemical performance LiFePO4/C composite cathode material with an ultrasonic-intensified micro-impinging jetting reactor.

    Science.gov (United States)

    Dong, Bin; Huang, Xiani; Yang, Xiaogang; Li, Guang; Xia, Lan; Chen, George

    2017-11-01

    A joint chemical reactor system referred to as an ultrasonic-intensified micro-impinging jetting reactor (UIJR), which possesses the feature of fast micro-mixing, was proposed and has been employed for rapid preparation of FePO 4 particles that are amalgamated by nanoscale primary crystals. As one of the important precursors for the fabrication of lithium iron phosphate cathode, the properties of FePO 4 nano particles significantly affect the performance of the lithium iron phosphate cathode. Thus, the effects of joint use of impinging stream and ultrasonic irradiation on the formation of mesoporous structure of FePO 4 nano precursor particles and the electrochemical properties of amalgamated LiFePO 4 /C have been investigated. Additionally, the effects of the reactant concentration (C=0.5, 1.0 and 1.5molL -1 ), and volumetric flow rate (V=17.15, 51.44, and 85.74mLmin -1 ) on synthesis of FePO 4 ·2H 2 O nucleus have been studied when the impinging jetting reactor (IJR) and UIJR are to operate in nonsubmerged mode. It was affirmed from the experiments that the FePO 4 nano precursor particles prepared using UIJR have well-formed mesoporous structures with the primary crystal size of 44.6nm, an average pore size of 15.2nm, and a specific surface area of 134.54m 2 g -1 when the reactant concentration and volumetric flow rate are 1.0molL -1 and 85.74mLmin -1 respectively. The amalgamated LiFePO 4 /C composites can deliver good electrochemical performance with discharge capacities of 156.7mAhg -1 at 0.1C, and exhibit 138.0mAhg -1 after 100 cycles at 0.5C, which is 95.3% of the initial discharge capacity. Copyright © 2017. Published by Elsevier B.V.

  9. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  10. Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

    The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

  11. 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)

  12. Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

    International Nuclear Information System (INIS)

    Dawahra, S.; Khattab, K.

    2011-01-01

    Highlights: → The MCNP4C code was used to calculate the power distribution in 3-D geometry in the MNSR reactor. → The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. → The minimum power was found in the fuel ring number 9 and was 79.9 W. → The total power in the total fuel rods was 30.9 kW. - Abstract: The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.

  13. Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

    Science.gov (United States)

    Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

    2017-12-01

    The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

  14. A new safety approach in the design of fast reactors

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

    1987-01-01

    A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

  15. Generation IV reactors: international projects

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

    2003-01-01

    Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

  16. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

  17. On alteration of reactor installation (additional installation of No.3 and No.4 plants in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc.)

    International Nuclear Information System (INIS)

    1985-01-01

    The Nuclear Safty Commission sent the reply to the Minister of International Trade and Industry on October 4, 1984, on this matter after having received the report from the Committee on Examination of Nuclear Reactor Safety and carried out the deliberation. It was judged that the applicant has the technical capability required for installing and operating these reactor facilities. Also it was judged that on the safety after these reactor plants are installed, there is no obstacle in the prevention of disaster due to contaminated substances and reactors. The policy of the investigation and deliberation is reported. The contents of the investigation and deliberation are the condition of location such as site, geological features and ground, earthquake, weather, hydraulic problem and social environments, the safety design of reactor facilities, the evaluation of radiation exposure dose in normal operation, the analysis of abnormal transient change in operation, accident analysis and the evaluation of location. (Kako, I.)

  18. Practical course on reactor instrumentation

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2004-06-01

    This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

  19. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  20. The Simulator Development for RDE Reactor

    Science.gov (United States)

    Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

  1. Repairing liner of the reactor

    International Nuclear Information System (INIS)

    Aguilar H, F.

    2001-07-01

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  3. Present status and future perspective of research and test reactors in JAERI

    International Nuclear Information System (INIS)

    Baba, Osamu; Kaieda, Keisuke

    1999-01-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  4. Present status and future perspective of research and test reactors in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

    1999-08-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  5. Three-phase packed bed reactor with an evaporating solvent—I. Experimental: the hydrogenation of 2,4,6-trinitrotoluene in methanol

    NARCIS (Netherlands)

    van Gelder, K.B.; Damhof, J.K.; Kroijenga, P.J.; Westerterp, K.R.

    1990-01-01

    In this paper we present experimental data on the three-phase hydrogenation of 2,4,6-trinitrotoluene (TNT) to triaminotoluene. The experiments are performed in a cocurrent upflow packed bed reactor. Methanol is used as an evaporating solvent. The influence of the main operating parameters, the

  6. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  7. Legal claims against Belgian reactors?; Rechtsmittel gegen belgische Reaktoren?

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian [CONLAR Consulting on Nuclear Law and Regulation, Leipzig (Germany)

    2016-06-15

    The Belgian reactors Tihange 2 and Doel 3 have been restarted in November 2015 after the problem of hydrogen flakes in the reactor pressure vessels had been investigated. The permission to restart has been the object both of critical statements by the German Federal Ministry of the Environment (BMUB) and of lawsuits filed with Belgian law courts by a group of German municipalities led by the city of Aachen and by the Land North-Rhine-Westphalia. According to a general principle of the law of nations, a state is not permitted to operate installations near its border, which cause significant environmental damage in a neighbouring state. However, it is not quite clear how this principle applies to the issue of potential accidents of nuclear power plants. According to the author, a tangible threat of an accident is required; mere doubts and concerns about the extent of safety margins are not sufficient.

  8. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  9. Commissioning of research reactors. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

  10. Permeated defect detecting test method and device in reactor

    International Nuclear Information System (INIS)

    Sakurai, Yoshishige.

    1996-01-01

    The present invention provides a method of and a device capable of performing a test for entire inner surfaces of the reactor upon periodical inspection of a BWR type reactor while sufficiently taking countermeasures for radiation rays into consideration. Namely, the present invention comprises following steps. (1) A provisional step for taking a shroud head of a reactor core shroud and incore structural components above and below the shroud out of the reactor, discharging reactor water and water tightly closing openings such as reactor wall perforation holes, (2) a pretreatment step for washing exposed inner surfaces of the reactor and peeling deteriorated materials, (3) a first drying step for drying portions washed and peeled in the step (2), (4) a permeation step for applying a permeation liquid of a defect detecting medium on the exposed inner surfaces of the reactor, (5) a permeation liquid removing step for removing the an excess permeation liquid in the step (4), (6) a second drying step for drying corresponding portions after performing the step (5), and (7) a flaw detecting step for optically observing the corresponding portions after performing the step (6) and detecting flaws. (I.S.)

  11. 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.)

  12. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  13. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  14. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  15. Status of the spent fuel in the reactor buildings of Fukushima Daiichi 1–4

    Energy Technology Data Exchange (ETDEWEB)

    Jäckel, Bernd S., E-mail: bernd.jaeckel@psi.ch

    2015-03-15

    The ratios of the radionuclides Cs-134g and Cs-137 deduced from measurements of liquid samples from the spent fuel pools in Fukushima Daiichi 1–4 are used to interpret the status of the spent fuel assemblies in the pools of the damaged reactor buildings. The different natures of the production of Cs-134g (neutron capture product of Cs-133) and Cs-137 (cumulative fission product from mass chain 137) and the different half-lives (2.06 years and 30.17 years respectively) require a complicated calculation of the mass and activity of the two nuclides. These masses are depending on the local burn up of the fuel, the burn up history and the radioactive decay. Calculation of the neutron capture product Cs-134g is particularly complicated, because the production of Cs-133 (stable cumulative fission product from mass chain 133) has to be taken into account. The neutron capture cross section for Cs-133 for thermal neutrons is well known, but the energy spectrum of the neutrons in a reactor includes higher energies according to the degree of moderation. Therefore the cross section was fitted from a gamma scan of spent fuel rods in a hot cell. The method of the calculation of the nuclide activities and the interpretation of the gamma measurements of the spent fuel pool samples from Fukushima Daiichi 1–4 are described in detail. It could be shown that at most only very minor mechanical damage of some spent fuel elements occurred during the accident and the later phase of the clearing work.

  16. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2000-01-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  17. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  18. Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

    Directory of Open Access Journals (Sweden)

    Siniša Šadek

    2010-01-01

    Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

  19. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

    2008-01-01

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  20. Progress of the decommissioning process of Musashi Institute of Technology reactor (4)

    International Nuclear Information System (INIS)

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

    2012-01-01

    The research reactor of Tokyo City University Atomic Energy Research Laboratory (Musashi Institute of Technology reactor) is zirconium-moderated water-cooled solid homogeneous type (TRIGA-II type), and its maximum heat output is 100 kW. It got into the first critical state in January 1963, and since then, it has mainly contributed to education and training for upgrading nuclear engineers, radioactivation analysis and reactor physics, and medical researches, as the joint usage research facilities across Japan. Then, after a long-term suspension, the university submitted the file in 2004 to the Ministry of Education, Culture, Sports, Science and Technology on the dismantling for the purpose of facility abolishment. Through the procedure of submitting a decommissioning plan, it was approved. Furthermore, in order to perform the function stop of the disposal facilities of liquid waste, application for change authorization for the decommissioning plan was submitted and approved. Regarding the progress of the decommissioning plan, the dismantling and removal of waste facilities for liquid waste and solid waste was carried out in FY2011 without any trouble. This paper explains this progress and future work plans. (A.O.)

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