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Sample records for babcock and wilcox lpr reactor

  1. Standard technical specifications for Babcock and Wilcox pressurized water reactors. Revision 4. Technical report

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors (BandW-STS) is a generic document prepared by the U.S. NRC for use in the licensing process. The BandW-STS provide applicants with model specifications to be used in formulation plant-specific technical specifications required by 10 CFR Part 50, Section 50.36, which set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection to the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  2. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls

  3. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  4. Standard Technical Specifications, Babcock and Wilcox Plants

    International Nuclear Information System (INIS)

    This NUREG contains improved Standard Technical Specifications (STS) for Babcock and Wilcox (B ampersand W) plants and documents the positions of the Nuclear Regulatory Commission (NRC) based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council. The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for developing improved plant-specific technical specifications by individual nuclear power plant licensees. This volume contains sections 3.4--3.9 which cover: Reactor coolant systems, emergency core cooling systems, containment systems, plant systems, electrical power systems, refueling operations

  5. Thermal-hydraulic research plan for Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    This document presents a plan for thermal-hydraulic research for Babcock and Wilcox designed reactor systems. It describes the technical issues, regulatory needs, and the research necessary to address these needs. The plan also discusses the relationship between current and proposed research, and provides a tentative schedule to complete the required work

  6. Production of leu high density fuels at Babcock and Wilcox

    International Nuclear Information System (INIS)

    A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

  7. Babcock and Wilcox Canada steam generators past, present and future

    International Nuclear Information System (INIS)

    The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

  8. LWRWIMS analysis of Babcock and Wilcox LWR fuel storage experiments

    International Nuclear Information System (INIS)

    The report describes very briefly an analysis of a series of critical experiments made by Babcock and Wilcox to study the relative importance on fuel storage reactivity of assembly spacing and various types of absorber. LWRWIMS in its standard design mode of calculation was used for the analysis. The results demonstrate that even the simplest options in LWRWIMS produce eigenvalues which are a very useful check of the Monte Carlo calculations normally made for criticality clearances. An appendix examines some of the eigenvalue trends in more detail. (author)

  9. Safety Evaluation Report related to Babcock and Wilcox Owners Group Plant Reassessment Program

    International Nuclear Information System (INIS)

    Supplement 1 to the ''Safety Evaluation Report (SER) Related to the Babcock and Wilcox Owners Group (BWOG's) Plant Reassessment Program'' has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). This supplement contains the NRC staff's evaluation of the BWOG reassessment of the integrated control system/non-nuclear instrumentation system, the emergency feedwater initiation and control system, reactor trip initiating events, several additional open items identified in the SER, and BWOG comments on the SER

  10. Radioactive waste shipments to Hanford retrievable storage from Babcock and Wilcox, Leechburg, Pennsylvania

    International Nuclear Information System (INIS)

    This report characterizes, as far as possible, the solid radioactive wastes generated by Babcock and Wilcox's Park Township Plutonium Facility near Leechburg, Pennsylvania that were sent to retrievable storage at the Hanford Site. Solid waste as defined in this document is any containerized or self-contained material that has been declared waste. The objective is a description of characteristics of solid wastes that are or will be managed by the Restoration and Upgrades Program; gaseous or liquid effluents are discussed only at a summary level This characterization is of particular interest in the planning of transuranic (TRU) waste retrieval operations, including the Waste Receiving and Processing (WRAP) Facility, because Babcock and Wilcox generated greater than 2.5 percent of the total volume of TRU waste currently stored at the Hanford Site

  11. Safety evaluation report related to Babcock and Wilcox Owners Group Plant Reassessment Program: [Final report

    International Nuclear Information System (INIS)

    After the accident of Three Mile Island, Unit 2, nuclear power plant owners made a number of improvements to their nuclear facilities. Despite these improvements, the US Nuclear Regulatory Commission (NRC) staff is concerned that the number and complexity of events at Babcock and Wilcox (B and W) nuclear plants have not decreased as expected. This concern was reinforced by the June 9, 1985 total-loss-of-feedwater event at Davis-Besse Nuclear Power Station and the December 26, 1985 overcooling transient at Rancho Seco Nuclear Generating Station. By letter dated January 24, 1986, the Executive Director for Operations (EDO) informed the Chairman of the B and W Owners Group (BWOG) that a number of recent events at B and W-designed reactors have led the NRC staff to conclude that the basic requirements for B and W reactors need to be reexamined. In its February 13, 1986 response to the EDO's letter, the BWOG committed to lead an effort to define concerns relative to reducing the frequency of reactor trips and the complexity of post-trip response in B and W plants. The BWOG submitted a description of the B and W program entitled ''Safety and Performance Improvement Program'' (BAW-1919) on May 15, 1986. Five revisions to BAW-1919 have also been submitted. The NRC staff has reviewed BAW-1919 and its revisions and presents its evaluation in this report. 2 figs., 34 tabs

  12. Comparison of licensing activities for operating plants designed by Babcock and Wilcox

    International Nuclear Information System (INIS)

    This report provides a comparison of a number of licensing activities for the operating Babcock and Wilcox (B and W) plants with emphasis on Rancho Seco. The factors selected were a comparison of staff resources expended in FY84, active licensing action reviews, implementation of NUREG-0737 modifications, exemptions to regulations, SALP reports, enforcement actions, and Licensee Event Reports (LERs). The eight licensed operating plants examined are as follows: Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3, Davis Besse, Oconee Units 1, 2, and 3, Rancho Seco, and Three Mile Island Unit 1 (TMI-1)

  13. Babcock and Wilcox Owners' Group program: Trip reduction and transient response improvement

    International Nuclear Information System (INIS)

    In 1985, the average trip frequency for the industry was 4.3 trips per plant per year while Babcock ampersand Wilcox (B ampersand W)-designed plants had 4.5 trips. In early 1986, the B ampersand W Owners' Group (B ampersand WOG) established goals to reduce trip frequency and improve posttrip transient response. Through the recommendations of the B ampersand WOG Trip Reduction and Transient Response Improvement Program (TR/TRIP) and other utility initiatives, the trip frequency for the B ampersand WOG plants has been on a progressive downward trend and has been consistently below the industry average since 1986. The successful results in trip reduction for the B ampersand WOG plants are shown. The B ampersand WOG has implemented several programs that have resulted in fewer trips per plant. This success can be attributed to the following: (1) a comprehensive program to evaluate each trip and transient for root-cause determination, define corrective actions, share information, and peer reviews; (2) a broad program to review systems and components that contribute to trips and transients, identify specific recommendations to correct deficiencies, utility commitment to implementation, conduct internal monitoring and indirectly exert peer pressure; (3) an awareness of the goals at all levels in the organization coupled with strong executive-level involvement; and (4) timely implementation of recommendations

  14. An aerial radiological survey of the Babcock and Wilcox Nuclear Facilities and surrounding area, Lynchburg, Virginia

    International Nuclear Information System (INIS)

    An aerial radiological survey was conducted from July 18 through July 25, 1988, over a 41-square-kilometer (16-square-mile) area surrounding the Babcock and Wilcox nuclear facilities located near Lynchburg, Virginia. The survey was conducted at a nominal altitude of 61 meters (200 feet) with line spacings of 91 meters (300 feet). A contour map of the terrestrial gamma exposure rate extrapolated to 1 meter above ground level (AGL) was prepared and overlaid on an aerial photograph. The terrestrial exposure rates varied from 8 to 12 microroentgens per hour (μR/h). A search of the data for man-made radiation sources revealed the presence of three areas of high count rates in the survey area. Spectra accumulated over the main plant showed the presence of cobalt-60 (60Co) and cesium-137 (137Cs). A second area near the main plant indicated the presence of uranium-235 (235U). Protactinium-234m (234mPa) and 60Co Were detected over a building to the east of the main plant. Soil samples and pressurized ion chamber measurements were obtained at four locations within the survey boundaries in support of the aerial data

  15. Standard technical specifications: Babcock and Wilcox plants. Volume 3, Revision 1: Bases (Sections 3.4--3.9)

    International Nuclear Information System (INIS)

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for Babcock and Wilcox Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS

  16. Comparison of implementation of selected TMI action plan requirements on operating plants designed by Babcock and Wilcox

    International Nuclear Information System (INIS)

    This report provides the results of a study conducted by the US Nuclear Regulatory Commission staff to compare the degree to which eight Babcock and Wilcox (B and W) designed licensed nuclear power plants have complied with the requirements in NUREG-0737, Clarification of TMI Action Plan Requirements. The eight licensed operating plants examined are as follows: Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3, Davis Besse, Oconee Units 1, 2, and 3, Rancho Seco, and Three Mile Island Unit 1 (TMI-1). The purpose of this audit was to establish the progress of the TMI-1 licensee, General Public Utilities (GPU) Nuclear Corporation, in completing the long-term requirements in NUREG-0737 relative to the other B and W licensees examined

  17. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    International Nuclear Information System (INIS)

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B ampersand W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B ampersand W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions

  18. Babcock and Wilcox Safety Anaysis Report (B-SAR-205). Volume 1

    International Nuclear Information System (INIS)

    The design of the BW-205 standard reactor with a plant output of 1295 and 1200 MW(e) is described. The reactor is arranged in two closed coolant loops connected in parallel to the reactor vessel, and is controlled by a coordinated combination of chemical shim and mechanical control rods. The coolant serves as a neutron moderator, reflector, and solvent for the soluble boron used in chemical shim reactivity control. The fuel elements consist of slightly enriched UO2 pellets enclosed in zircaloy tubes

  19. Compact Process Development at Babcock & Wilcox

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Jeffrey Phillips

    2012-03-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of compaction trials have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel at packing fractions exceeding 46% by volume. Results from these trials are included. The scale-up effort is nearing completion with the process installed and operable using nuclear fuel materials. Final process testing is in progress to certify the process for manufacture of qualification test fuel compacts in 2012.

  20. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed.

  1. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed

  2. Benchmarking of flowtran with Mark-22 mockup flow excursion test data from Babcock ampersand Wilcox

    International Nuclear Information System (INIS)

    Version 16.2 of the FLOWTRAN code with a Savannah River Site (SRS) working criterion (St=0.00455) for the onset of significant void (OSV) was benchmarked against power and flow excursion data derived from tests at the Babcock ampersand Wilcox Alliance Research Center test facility. This document presents analyses which show that FLOWTRAN accurately predicts the mockup test assembly thermal-hydraulic behavior during the steady state and LOCA transient conditions, and that FLOWTRAN with a Savannah River Site (SRS) working limits criterion (St=0.00455) conservatively predicts the OFI power

  3. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

  4. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.

  5. Estimate of airborne release of plutonium from Babcock and Wilcox plant as a result of severe wind hazard and earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, J.; Schwendiman, L.C.; Ayer, J.E.

    1978-10-01

    As part of an interdisciplinary study to evaluate the potential radiological consequences of wind hazard and earthquake upon existing commercial mixed oxide fuel fabrication plants, the potential mass airborne releases of plutonium (source terms) from such events are estimated. The estimated souce terms are based upon the fraction of enclosures damaged to three levels of severity (crush, puncture penetrate, and loss of external filter, in order of decreasing severity), called damage ratio, and the airborne release if all enclosures suffered that level of damage. The discussion of damage scenarios and source terms is divided into wind hazard and earthquake scenarios in order of increasing severity. The largest airborne releases from the building were for cases involving the catastrophic collapse of the roof over the major production areas--wind hazard at 110 mph and earthquakes with peak ground accelerations of 0.20 to 0.29 g. Wind hazards at higher air velocities and earthquakes with higher ground acceleration do not result in significantly greater source terms. The source terms were calculated as additional mass of respirable particles released with time up to 4 days; and, under these assumptions, approximately 98% of the mass of material of concern is made airborne from 2 h to 4 days after the event. The overall building source terms from the damage scenarios evaluated are shown in a table. The contribution of individual areas to the overall building source term is presented in order of increasing severity for wind hazard and earthquake.

  6. GERD (Gastroesophageal Reflux) and LPR (Laryngopharyngeal Reflux)

    Science.gov (United States)

    ... Meeting Calendar Find an ENT Doctor Near You GERD and LPR GERD and LPR Patient Health Information ... relations staff at newsroom@entnet.org . What is GERD? Gastroesophageal reflux disease, often referred to as GERD, ...

  7. Replacement steam generators for pressurized water reactors

    International Nuclear Information System (INIS)

    Babcock and Wilcox Canada has developed an Advanced Series steam generator for PWR Systems. This design incorporates all of the features that have contributed to the successful CANDU steam generator performance. This paper presents an overview of the design features and how the overall design relates to the requirements of a PWR reactor system

  8. Reactor internals design/analysis for normal, upset, and faulted conditions

    International Nuclear Information System (INIS)

    The analytical procedures used by Babcock and Wilcox to demonstrate the structural integrity of the 205-FA reactor internals are described. Analytical results are presented and compared to ASME Code allowable limits for Normal, Upset, and Faulted conditions. The particular faulted condition considered is a simultaneous loss-of-coolant accident and safe shutdown earthquake. The operating basis earthquake is addressed as an Upset condition

  9. Report of the Bulletins and Orders Task Force. Volume II. Appendices

    International Nuclear Information System (INIS)

    Appendices include: Office of Inspection and Enforcement bulletins; NRR status report on feedwater transients in BWR plants; orders on Babcock and Wilcox Company plants; letters lifting orders; letters issuing auxiliary feedwater system requirements; letter to licensees of all operating reactors, dated October 30, 1979 concerning short-term lessons learned requirements; and letters approving guidelines for preparation of small-break LOCA operating procedures

  10. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    OpenAIRE

    Karak, Bidya Binay; Cameron, Robert

    2016-01-01

    The key elements of the Babcock-Leighton dynamo are the generation of poloidal field through the decay of tilted bipolar active regions and the generation of toroidal field through the observed differential rotation. There are two classes of Babcock-Leighton models: flux transport dynamos where an equatorward flow at the bottom of the convection zone (CZ) causes the equatorial propagation of the butterfly wings, and dynamo waves where the radial shear and the $\\alpha$ effect act in conjunctio...

  11. Defects in the peripheral taste structure and function in the MRL/lpr mouse model of autoimmune disease.

    Directory of Open Access Journals (Sweden)

    Agnes Kim

    Full Text Available While our understanding of the molecular and cellular aspects of taste reception and signaling continues to improve, the aberrations in these processes that lead to taste dysfunction remain largely unexplored. Abnormalities in taste can develop in a variety of diseases, including infections and autoimmune disorders. In this study, we used a mouse model of autoimmune disease to investigate the underlying mechanisms of taste disorders. MRL/MpJ-Fas(lpr/J (MRL/lpr mice develop a systemic autoimmunity with phenotypic similarities to human systemic lupus erythematosus and Sjögren's syndrome. Our results show that the taste tissues of MRL/lpr mice exhibit characteristics of inflammation, including infiltration of T lymphocytes and elevated levels of some inflammatory cytokines. Histological studies reveal that the taste buds of MRL/lpr mice are smaller than those of wild-type congenic control (MRL/+/+ mice. 5-Bromo-2'-deoxyuridine (BrdU pulse-chase experiments show that fewer BrdU-labeled cells enter the taste buds of MRL/lpr mice, suggesting an inhibition of taste cell renewal. Real-time RT-PCR analyses show that mRNA levels of several type II taste cell markers are lower in MRL/lpr mice. Immunohistochemical analyses confirm a significant reduction in the number of gustducin-positive taste receptor cells in the taste buds of MRL/lpr mice. Furthermore, MRL/lpr mice exhibit reduced gustatory nerve responses to the bitter compound quinine and the sweet compound saccharin and reduced behavioral responses to bitter, sweet, and umami taste substances compared with controls. In contrast, their responses to salty and sour compounds are comparable to those of control mice in both nerve recording and behavioral experiments. Together, our results suggest that type II taste receptor cells, which are essential for bitter, sweet, and umami taste reception and signaling, are selectively affected in MRL/lpr mice, a model for autoimmune disease with chronic

  12. The interactive effect of MAOA-LPR genotype and childhood physical neglect on aggressive behaviors in Italian male prisoners

    OpenAIRE

    Gorodetsky, Elena; Bevilacqua, Laura; Carli, Vladimir; Sarchiapone, Marco; Roy, Alec; Goldman, David; Enoch, Mary-Anne

    2014-01-01

    Aggressive disorders are moderately heritable; therefore, identification of genetic influences is important. The X-linked MAOA gene, encoding the MAOA enzyme, has a functional 30bp repeat polymorphism in the promoter region (MAOA-LPR) that has been shown to influence aggression. Childhood trauma is a known risk factor for numerous psychopathologies in adulthood including aggressive behaviors. We investigated the interactive effect of MAOA-LPR genotype and a history of childhood trauma in pred...

  13. Preliminary study of uranium favorability of the Wilcox and Claiborne Groups (Eocene) in Texas

    Energy Technology Data Exchange (ETDEWEB)

    Wilbert, W.P.; Templain, C.J.

    1978-01-01

    Rocks of the Wilcox and Claiborne Groups crop out in the Texas Gulf Coastal Plain and are represented by a series of sands and shales which reflect oscillation of the strandline. The Wilcox Group (lower Eocene), usually undifferentiated in Texas, consists of very fine sands and clays and abundant lignite. The Claiborne Group (middle Eocene) comprises, in ascending order, Carrizo Sand, Reklaw Formation (clay), Queen City Sand, Weches Formation (clay), Sparta Sand, Cook Mountain Formation (clay), and Yegua Formation (sand). Fluvial systems of the Wilcox and Claiborne Groups exist in east Texas and trend perpendicular to the present coastline. In central Texas, sand bodies are parallel to the present coastline and are strand-plain, barrier-bar systems. Since the time of deposition of the Queen City Sand, a significant fluvial sand buildup occurred in the area of the present Rio Grande embayment where the marine clays pinch out. Known occurrences of mineral matter in the Wilcox and Claiborne (up to the Yegua) are limited to lignite (particularly in the Wilcox), cannel coal in the upper Claiborne, and hydrocarbons throughout. No uranium mineralization is known, and no uranium is likely to be discovered in the Claiborne and Wilcox. Approximately 50 surface samples and many gamma-ray logs showed no significant anomalies. The sands are very good potential host rocks, but no uranium source was discovered. During deposition of the Wilcox and Claiborne Groups, there was no volcanism to serve as a source of uranium (as with the prolific occurrences in the younger rocks of south Texas); also, Precambrian crystalline rocks in the Llano uplift were not exposed.

  14. Susceptible cytotoxicity to ultraviolet B light in fibroblasts and keratinocytes cultured from autoimmune-prone MRL/Mp-lpr/lpr mice

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, F.; Lyon, M.B.; Norris, D.A. (Univ. of Colorado School of Medicine, Denver (USA))

    1989-09-01

    The MRL/Mp-lpr/lpr (MRL/l) mouse is an autoimmune model of spontaneous lupus erythematosus (LE), in addition to lupus nephritis. In order to better understand the mechanisms of photosensitivity in LE, in vitro photocytotoxicity was examined by using fibroblasts and keratinocytes cultured from MRL/l mice, control MRL/Mp- +/+ (MRL/n) mice, and normal BALB/c mice. A colorimetric 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyl tetrazolium bromide assay and the acridine orange/ethidium bromide assay were used for determination of cytotoxicity. Fibroblasts cultured from newborn MRL/l mice showed higher susceptibility to single ultraviolet light B (UVB) light irradiation at a dose of 100-500 mJ than those from MRL/n, F1 hybrid of (MRL/l x MRL/n mice), and BALB/c mice. However, the susceptibility to UVB was not observed in young (1-month-old) and adult (4-month-old) MRL/l mice. UVA light irradiation was not cytotoxic. Keratinocytes cultured from MRL mice showed lower cytotoxicity to UVB irradiation than fibroblasts cultured. However, keratinocytes from newborn MRL/l mice showed higher cytotoxicity to 50 mJ UVB irradiation than cells from MRL/n mice. Syngeneic or allogeneic sera augmented UVB-induced cytotoxicity of fibroblasts cultured. UVB irradiation of spleen cells induced no significant difference of cytotoxicity between MRL/l and MRL/n mice. Based on the results of F1 hybrid of (MRL/l x MRL/n) mice, the susceptibility seemed to be associated with autoimmune traits and to be regulated by genetical background.

  15. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    CERN Document Server

    Karak, Bidya Binay

    2016-01-01

    The key elements of the Babcock-Leighton dynamo are the generation of poloidal field through the decay of tilted bipolar active regions and the generation of toroidal field through the observed differential rotation. There are two classes of Babcock-Leighton models: flux transport dynamos where an equatorward flow at the bottom of the convection zone (CZ) is responsible for the equatorial propagation of the butterfly wings, and dynamo waves where the radial gradient of differential rotation and the $\\alpha$ effect act in conjunction to produce the equatorial propagation. Here we investigate the role of downward magnetic pumping near the surface using a kinematic Babcock-Leighton model. We find that the pumping causes the poloidal field to become predominately radial in the near-surface shear layer. This allows the negative radial shear in the near-surface layer to effectively act on the radial field to produce a toroidal field. Consequently, we observe a clear equatorward migration of the toroidal field at lo...

  16. Mycobacterial Metabolic Syndrome: LprG and Rv1410 Regulate Triacylglyceride Levels, Growth Rate and Virulence in Mycobacterium tuberculosis.

    Directory of Open Access Journals (Sweden)

    Amanda J Martinot

    2016-01-01

    Full Text Available Mycobacterium tuberculosis (Mtb mutants lacking rv1411c, which encodes the lipoprotein LprG, and rv1410c, which encodes a putative efflux pump, are dramatically attenuated for growth in mice. Here we show that loss of LprG-Rv1410 in Mtb leads to intracellular triacylglyceride (TAG accumulation, and overexpression of the locus increases the levels of TAG in the culture medium, demonstrating a role of this locus in TAG transport. LprG binds TAG within a large hydrophobic cleft and is sufficient to transfer TAG from donor to acceptor membranes. Further, LprG-Rv1410 is critical for broadly regulating bacterial growth and metabolism in vitro during carbon restriction and in vivo during infection of mice. The growth defect in mice is due to disrupted bacterial metabolism and occurs independently of key immune regulators. The in vivo essentiality of this locus suggests that this export system and other regulators of metabolism should be considered as targets for novel therapeutics.

  17. Mycobacterial Metabolic Syndrome: LprG and Rv1410 Regulate Triacylglyceride Levels, Growth Rate and Virulence in Mycobacterium tuberculosis.

    Science.gov (United States)

    Martinot, Amanda J; Farrow, Mary; Bai, Lu; Layre, Emilie; Cheng, Tan-Yun; Tsai, Jennifer H; Iqbal, Jahangir; Annand, John W; Sullivan, Zuri A; Hussain, M Mahmood; Sacchettini, James; Moody, D Branch; Seeliger, Jessica C; Rubin, Eric J

    2016-01-01

    Mycobacterium tuberculosis (Mtb) mutants lacking rv1411c, which encodes the lipoprotein LprG, and rv1410c, which encodes a putative efflux pump, are dramatically attenuated for growth in mice. Here we show that loss of LprG-Rv1410 in Mtb leads to intracellular triacylglyceride (TAG) accumulation, and overexpression of the locus increases the levels of TAG in the culture medium, demonstrating a role of this locus in TAG transport. LprG binds TAG within a large hydrophobic cleft and is sufficient to transfer TAG from donor to acceptor membranes. Further, LprG-Rv1410 is critical for broadly regulating bacterial growth and metabolism in vitro during carbon restriction and in vivo during infection of mice. The growth defect in mice is due to disrupted bacterial metabolism and occurs independently of key immune regulators. The in vivo essentiality of this locus suggests that this export system and other regulators of metabolism should be considered as targets for novel therapeutics. PMID:26751071

  18. A functional polymorphism in the MAOA gene promoter (MAOA-LPR) predicts central dopamine function and body mass index.

    Science.gov (United States)

    Ducci, F; Newman, T K; Funt, S; Brown, G L; Virkkunen, M; Goldman, D

    2006-09-01

    Variation in brain monoaminergic activity is heritable and modulates risk of alcoholism and other addictions, as well as food intake and energy expenditure. Monoamine oxidase A deaminates the monoamine neurotransmitters serotonin, dopamine (DA), and noradrenaline. The monoamine oxidase A (MAOA) gene (Xp11.5) contains a length polymorphism in its promoter region (MAOA-LPR) that putatively affects transcriptional efficiency. Our goals were to test (1) whether MAOA-LPR contributes to interindividual variation in monoamine activity, assessed using levels of cerebrospinal fluid (CSF) monoamine metabolites; and (2) whether MAOA-LPR genotype influences alcoholism and/or body mass index (BMI). Male, unrelated criminal alcoholics (N=278) and controls (N=227) were collected from a homogeneous Finnish source population. CSF concentration of 5-hydroxyindoleacetic acid (5-HIAA), homovanillic acid (HVA), and 3-methoxy-4-hydroxyphenylglycol (MHPG) were available from 208 participants. Single allele, hemizygous genotypes were grouped according to inferred effect of the MAOA alleles on transcriptional activity. MAOA-LPR genotypes had a significant effect on CSF HVA concentration (P=0.01) but explained only 3% of the total variance. There was a detectable but nonsignificant genotype effect on 5-HIAA and no effects on MHPG. Specifically, the genotype conferring high MAOA activity was associated with lower HVA levels in both alcoholics and controls, a finding that persisted after accounting for the potential confounds of alcoholism, BMI, height, and smoking. MAOA-LPR genotype predicted BMI (P<0.005), with the high-activity genotype being associated with lower BMI. MAOA-LPR genotypes were not associated with alcoholism or related psychiatric phenotypes in this data set. Our results suggest that MAOA-LPR allelic variation modulates DA turnover in the CNS, but does so in a manner contrary to our prior expectation that alleles conferring high activity would predict higher HVA levels in

  19. Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    In a memo dated 19 August 1993, Secretary O'Leary assigned the Office of Environment, Safety and Health the primary responsibility to identify, characterize, and assess the safety, health, and environmental vulnerabilities of the DOE's existing storage conditions and facilities for the storage of irradiated reactor fuel and other reactor irradiated nuclear materials. This volume is divided into three major sections. Section 1 contains the Working Group Assessment Team reports on the following facilities: Hanford Site, INEL, SRS, Oak Ridge Site, West Valley Site, LANL, BNL, Sandia, General Atomics (San Diego), Babcock ampersand Wilcox (Lynchburg Technology Center), and ANL. Section 2 contains the Vulnerability Development Forms from most of these sites. Section 3 contains the documents used by the Working Group in implementing this initiative

  20. Wilcox Group Apparent Thickness, Gulf Coast (wlcxthkg)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Apparent Wilcox Group thickness maps are contoured from location and top information derived from the Petroleum Information (PI) Wells database. The Wilcox...

  1. Magnetic flux transport and the sun's dipole moment - New twists to the Babcock-Leighton model

    Science.gov (United States)

    Wang, Y.-M.; Sheeley, N. R., Jr.

    1991-01-01

    The mechanisms that give rise to the sun's large-scale poloidal magnetic field are explored in the framework of the Babcock-Leighton (BL) model. It is shown that there are in general two quite distinct contributions to the generation of the 'alpha effect': the first is associated with the axial tilts of the bipolar magnetic regions as they erupt at the surface, while the second arises through the interaction between diffusion and flow as the magnetic flux is dispersed over the surface. The general relationship between flux transport and the BL dynamo is discussed.

  2. White Paper on the Use of Team Calendars with the JIRA Issue Tracking System and Confluence Collaboration Tools for the xLPR Project

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Hilda B [ORNL; Williams, Paul T [ORNL; Bass, Bennett Richard [ORNL

    2012-09-01

    ORNL was tasked by xLPR project management to propose a team calendar for use within the xLPR consortium. Among various options that were considered, the approach judged by ORNL to best fit the needs of the xLPR project is presented in this document. The Atlassian Team Calendars plug-in used with the Confluence collaboration tool was recommended for several reasons, including the advantage that it provides for a tight integration between Confluence (found at https://xlpr.ornl.gov/wiki ) and xLPR s JIRA issue tracking system (found at https://xlpr.ornl.gov/jira ). This document is divided into two parts. The first part (Sections 1-6) consists of the white paper, which highlights some of the ways that Team Calendars can improve com mun ication between xLPR project managers, group leads, and team members when JIRA is applied for both issue tracking and change-management activities. Specific points emphasized herein are as follows: The Team Calendar application greatly enhances the added value that the JIRA and Confluence tools bring to the xLPR Project. The Team Calendar can improve com mun ication between xLPR project managers, group leads, and team members when JIRA is applied for both issue tracking and change-management activities. The Team Calendar works across different email tools such as Outlook 2011, Outlook 2010, Outlook 2007, Google Calendars and Mac s iCalendar to name a few. xLPR users can now access the wiki Confluence (with embedded Team Calendars) directly from JIRA without having to re-validate their login. The second part consists of an Annex (Section 7), which describes how users can subscribe to Team Calendars from different calendar applications. Specific instructions are given in the Annex that describe how to Import xLPR Team Calendar to Outlook Version Office 2010 Import xLPR Team Calendar to Outlook Version Office 2007 Subscribe to Team Calendar from Google Calendar The reader is directed to Section 4 for instructions on adding events to the

  3. Technical and economic studies of small reactors for supply of electricity and steam

    Energy Technology Data Exchange (ETDEWEB)

    Spiewak, I.; Klepper, O. H.; Fuller, L. C.

    1977-02-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination.

  4. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination

  5. U.S. Nuclear Regulatory Commission Extremely Low Probability of Rupture pilot study : xLPR framework model user's guide.

    Energy Technology Data Exchange (ETDEWEB)

    Kalinich, Donald A.; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01

    For the U.S. Nuclear Regulatory Commission (NRC) Extremely Low Probability of Rupture (xLPR) pilot study, Sandia National Laboratories (SNL) was tasked to develop and evaluate a probabilistic framework using a commercial software package for Version 1.0 of the xLPR Code. Version 1.0 of the xLPR code is focused assessing the probability of rupture due to primary water stress corrosion cracking in dissimilar metal welds in pressurizer surge nozzles. Future versions of this framework will expand the capabilities to other cracking mechanisms, and other piping systems for both pressurized water reactors and boiling water reactors. The goal of the pilot study project is to plan the xLPR framework transition from Version 1.0 to Version 2.0; hence the initial Version 1.0 framework and code development will be used to define the requirements for Version 2.0. The software documented in this report has been developed and tested solely for this purpose. This framework and demonstration problem will be used to evaluate the commercial software's capabilities and applicability for use in creating the final version of the xLPR framework. This report details the design, system requirements, and the steps necessary to use the commercial-code based xLPR framework developed by SNL.

  6. Interpretational Confounding Is Due to Misspecification, Not to Type of Indicator: Comment on Howell, Breivik, and Wilcox (2007)

    Science.gov (United States)

    Bollen, Kenneth A.

    2007-01-01

    R. D. Howell, E. Breivik, and J. B. Wilcox (2007) have argued that causal (formative) indicators are inherently subject to interpretational confounding. That is, they have argued that using causal (formative) indicators leads the empirical meaning of a latent variable to be other than that assigned to it by a researcher. Their critique of causal…

  7. Apparent Depth to the Wilcox Group, Gulf Coast (wlcxdpthg)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The depth to top of the Wilcox Group is contoured from location and top information derived from the Petroleum Information (PI) Wells database. The depth to Wilcox...

  8. Preliminary Design Concept for a Reactor-internal CRDM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later.

  9. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

  10. Development, analysis, and evaluation of a commercial software framework for the study of Extremely Low Probability of Rupture (xLPR) events at nuclear power plants.

    Energy Technology Data Exchange (ETDEWEB)

    Kalinich, Donald A.; Helton, Jon Craig; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01

    Sandia National Laboratories (SNL) participated in a Pilot Study to examine the process and requirements to create a software system to assess the extremely low probability of pipe rupture (xLPR) in nuclear power plants. This project was tasked to develop a prototype xLPR model leveraging existing fracture mechanics models and codes coupled with a commercial software framework to determine the framework, model, and architecture requirements appropriate for building a modular-based code. The xLPR pilot study was conducted to demonstrate the feasibility of the proposed developmental process and framework for a probabilistic code to address degradation mechanisms in piping system safety assessments. The pilot study includes a demonstration problem to assess the probability of rupture of DM pressurizer surge nozzle welds degraded by primary water stress-corrosion cracking (PWSCC). The pilot study was designed to define and develop the framework and model; then construct a prototype software system based on the proposed model. The second phase of the project will be a longer term program and code development effort focusing on the generic, primary piping integrity issues (xLPR code). The results and recommendations presented in this report will be used to help the U.S. Nuclear Regulatory Commission (NRC) define the requirements for the longer term program.

  11. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    Science.gov (United States)

    Karak, Bidya Binay; Cameron, Robert

    2016-05-01

    We investigate the role of downward magnetic pumping near the surface using a kinematic Babcock-Leighton model. We find that the pumping causes the poloidal field to become predominately radial in the near-surface shear layer. This allows the negative radial shear in the near-surface layer to effectively act on the radial field to produce a toroidal field. Consequently, we observe a clear equatorward migration of the toroidal field at low latitudes even when there is no meridional flow in the deep CZ. We show a case where the period of a dynamo wave solution is approximately 11 years. Flux transport models are also shown with periods close to 11 years. Both the dynamo wave and flux transport dynamo are thus able to reproduce some of the observed features of solar cycle. The main difference between the two types of dynamo is the value of $\\alpha$ required to produce dynamo action. In both types of dynamo, the surface meridional flow helps to advect and build the polar field in high latitudes, while in flux transport dynamo the equatorward flow near the bottom of CZ advects toroidal field to cause the equatorward migration in butterfly wings and this advection makes the dynamo easier by transporting strong toroidal field to low latitudes where $\\alpha$ effect works. Another conclusion of our study is that the magnetic pumping suppresses the diffusion of fields through the photospheric surface which helps to achieve the 11-year dynamo cycle at a moderately larger value of magnetic diffusivity than has previously been used.

  12. Deletion of microRNA-155 reduces autoantibody responses and alleviates lupus-like disease in the Fas(lpr) mouse

    NARCIS (Netherlands)

    Thai, To-Ha; Patterson, Heide Christine; Pham, Duc-Hung; Kis-Toth, Katalin; Kaminski, Denise A.; Tsokos, George C.

    2013-01-01

    MicroRNA-155 (miR-155) regulates antibody responses and subsequent B-cell effector functions to exogenous antigens. However, the role of miR-155 in systemic autoimmunity is not known. Using the death receptor deficient (Fas(lpr)) lupus-prone mouse, we show here that ablation of miR-155 reduced autoa

  13. Deletion of microRNA-155 reduces autoantibody responses and alleviates lupus-like disease in the Fas(lpr) mouse.

    Science.gov (United States)

    Thai, To-Ha; Patterson, Heide Christine; Pham, Duc-Hung; Kis-Toth, Katalin; Kaminski, Denise A; Tsokos, George C

    2013-12-10

    MicroRNA-155 (miR-155) regulates antibody responses and subsequent B-cell effector functions to exogenous antigens. However, the role of miR-155 in systemic autoimmunity is not known. Using the death receptor deficient (Fas(lpr)) lupus-prone mouse, we show here that ablation of miR-155 reduced autoantibody responses accompanied by a decrease in serum IgG but not IgM anti-dsDNA antibodies and a reduction of kidney inflammation. MiR-155 deletion in Fas(lpr) B cells restored the reduced SH2 domain-containing inositol 5'-phosphatase 1 to normal levels. In addition, coaggregation of the Fc γ receptor IIB with the B-cell receptor in miR-155(-/-)-Fas(lpr) B cells resulted in decreased ERK activation, proliferation, and production of switched antibodies compared with miR-155 sufficient Fas(lpr) B cells. Thus, by controlling the levels of SH2 domain-containing inositol 5'-phosphatase 1, miR-155 in part maintains an activation threshold that allows B cells to respond to antigens. PMID:24282294

  14. Piperlongumine alleviates lupus nephritis in MRL-Fas(lpr) mice by regulating the frequency of Th17 and regulatory T cells.

    Science.gov (United States)

    Yao, Lan; Chen, Hai-ping; Ma, Qing

    2014-09-01

    Recent data have shown that piperlongumine (PL), an important component of Piper longum fruits, is known to possess anti-inflammatory and vascular-protective activities. This study aimed to examine the therapeutic effects and underlying mechanisms of PL on lupus-prone MRL-Fas(lpr) mice. Female MRL-Fas(lpr) mice were intraperitoneally treated with PL (2.4 mg kg(-1) d(-1)) for 10 weeks, and the proteinuria level was biweekly monitored. After the mice were euthanized, serum biochemical parameters and renal damage were determined. Splenocytes of MRL-Fas(lpr) mice were isolated for in vitro study. Treatment of the mice with PL significantly attenuated the progression of proteinuria and glomerulonephritis. The improvement was accompanied by decreased serum levels of nephritogenic anti-dsDNA antibodies, IL-6, IL-17, IL-23 and TNF-α. Treatment of the mice with PL suppressed the frequency of Th17 cells and increased the regulatory T cells (Tregs). In vitro, the levels of IL-6, IL-17, IL-23 and TNF-α were significantly decreased in the cultures of splenocytes from PL-treated mice compared with those from vehicle-treated mice. In addition, PL treatment impeded activation of the JAK/STAT3 signaling in splenocytes. Of great important, the survival of MRL-Fas(lpr) mice were improved by PL treatment. In summary, PL effectively ameliorates lupus syndrome in MRL-Fas(lpr) mice by suppressing the pathogenic Th17 cells and increasing the Tregs as well as inhibiting activation of the JAK/STAT3 signaling pathway. This study sheds new light on the immune-modulatory role of PL. PMID:24837470

  15. A Babcock-Leighton solar dynamo model with multi-cellular meridional circulation in advection- and diffusion-dominated regimes

    CERN Document Server

    Belucz, Bernadett; Forgacs-Dajka, Emese

    2015-01-01

    Babcock-Leighton type solar dynamo models with single-celled meridional circulation are successful in reproducing many solar cycle features. Recent observations and theoretical models of meridional circulation do not indicate a single-celled flow pattern. We examine the role of complex multi-cellular circulation patterns in a Babcock-Leighton solar dynamo in advection- and diffusion-dominated regimes. We show from simulations that presence of a weak, second, high-latitude reverse cell speeds up the cycle and slightly enhances the poleward branch in butterfly diagram, whereas the presence of a second cell in depth reverses the tilt of butterfly wing to an anti-solar type. A butterfly diagram constructed from middle of convection zone yields a solar-like pattern, but this may be difficult to realize in the Sun because of magnetic buoyancy effects. Each of the above cases behaves similarly in higher and lower magnetic diffusivity regimes. However, our dynamo with a meridional circulation containing four cells in...

  16. Organic petrology and coalbed gas content, Wilcox Group (Paleocene-Eocene), northern Louisiana

    Science.gov (United States)

    Hackley, P.C.; Warwick, P.D.; Breland, F.C.

    2007-01-01

    Wilcox Group (Paleocene-Eocene) coal and carbonaceous shale samples collected from four coalbed methane test wells in northern Louisiana were characterized through an integrated analytical program. Organic petrographic analyses, gas desorption and adsorption isotherm measurements, and proximate-ultimate analyses were conducted to provide insight into conditions of peat deposition and the relationships between coal composition, rank, and coalbed gas storage characteristics. The results of petrographic analyses indicate that woody precursor materials were more abundant in stratigraphically higher coal zones in one of the CBM wells, consistent with progradation of a deltaic depositional system (Holly Springs delta complex) into the Gulf of Mexico during the Paleocene-Eocene. Comparison of petrographic analyses with gas desorption measurements suggests that there is not a direct relationship between coal type (sensu maceral composition) and coalbed gas storage. Moisture, as a function of coal rank (lignite-subbituminous A), exhibits an inverse relationship with measured gas content. This result may be due to higher moisture content competing for adsorption space with coalbed gas in shallower, lower rank samples. Shallower ( 600??m) coal samples containing less moisture range from under- to oversaturated with respect to their CH4 adsorption capacity.

  17. miR-155 Deficiency Ameliorates Autoimmune Inflammation of Systemic Lupus Erythematosus by Targeting S1pr1 in Faslpr/lpr Mice.

    Science.gov (United States)

    Xin, Qian; Li, Jiangxia; Dang, Jie; Bian, Xianli; Shan, Shan; Yuan, Jupeng; Qian, Yanyan; Liu, Zhaojian; Liu, Guangyi; Yuan, Qianqian; Liu, Na; Ma, Xiaochun; Gao, Fei; Gong, Yaoqin; Liu, Qiji

    2015-06-01

    MicroRNA-155 (miR-155) was previously found involved in the development of systemic lupus erythematosus (SLE) and other autoimmune diseases and the inflammatory response; however, the detailed mechanism of miR-155 in SLE is not fully understood. To explore the in vivo role of miR-155 in the pathogenesis of SLE, miR-155-deficient Fas(lpr/lpr) (miR-155(-/-)Fas(lpr/lpr)) mice were obtained by crossing miR-155(-/-) and Fas(lpr/lpr) mice. Clinical SLE features such as glomerulonephritis, autoantibody levels, and immune system cell populations were compared between miR-155(-/-)Fas(lpr/lpr) and Fas(lpr/lpr) mice. Microarray analysis, RT-PCR, Western blot, and luciferase reporter gene assay were used to identify the target gene of miR-155. miR-155(-/-)Fas(lpr/lpr) mice showed milder SLE clinical features than did Fas(lpr/lpr)mice. As compared with Fas(lpr/lpr) mice, miR-155(-/-)Fas(lpr/lpr) mice showed less deposition of total IgA, IgM, and IgG and less infiltration of inflammatory cells in the kidney. Moreover, the serum levels of IL-4 and IL-17a, secreted by Th2 and Th17 cells, were lower in miR-155(-/-)Fas(lpr/lpr) than Fas(lpr/lpr) mice; the CD4(+)/CD8(+) T cell ratio was restored in miR-155(-/-)Fas(lpr/lpr) mice as well. Sphingosine-1-phosphate receptor 1 (S1PR1) was found as a new target gene of miR-155 by in vitro and in vivo studies; its expression was decreased in SLE patients and Fas(lpr/lpr) mice. miR-155(-/-)Fas(lpr/lpr) mice are resistant to the development of SLE by the regulation of the target gene S1pr1. miR-155 might be a new target for therapeutic intervention in SLE. PMID:25911753

  18. The New US Public-Private Partnership to License and DeploySmall Modular Reactors, With Focus on The B and W mPower Reactor

    International Nuclear Information System (INIS)

    On December 16, 2011, The US Congress and the President approved new Fiscal Year 2012 funding for a Government - Industry cost shared program called 'Small Modular Reactor (SMR) Licensing Technical Support'. The new legislation appropriates $67 million in 2012 to provide licensing and first-of-a-kind engineering support for small modular reactor designs that can be deployed expeditiously. The legislation requires the Department of Energy to consider applications utilizing any small modular reactor technology. Competitive solicitations are likely to begin shortly and two or three SM R designs will be selected for U S Government support. Such support will likely accelerate deployment and operation of at least one such design by 2020. The Congressional language states that the Government portion of the program is expected to total $452 million over five years One of the candidates for this competition is the B and W mPower reactor being developed by Generation mPower, a company recently formed by the Babcock and Wilcox Company and Bechtel Power Corporation. This presentation will summarize the main features of this design, and explain why it meets the requirements for the Government program, and will be fully developed, licensed and deployed in the US within the next 8 years. Importantly, this design has many features that favor its introduction and use in smaller countries with critical needs for future electric generation capacity, with arid conditions that may require air cooled condensers, and with potential need for a desalination component of the new energy source. The relatively small capacity of the modules (e.g. 320 MWe for an initial two unit plant) will require much lower initial capital investment, as compared to the very large investment of $4 to $6 billion required for the newer 1100 to 1400 MWe plants now being constructed in China, France, Finland, Korea, the US, and the United Arab Emirates

  19. Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    Each Site Team, consisting of M ampersand O contractor and Operations Office personnel, performed data collection and identified ES ampersand H concerns relative to RINM storage by preparing responses to the detailed question set for each storage facility at the site. These responses formed the basis for the Site Team reports. These reports are contained in this volume and are from the following facilities: Hanford Site, Idaho National Engineering Laboratory Site, Savannah River Site, Oak Ridge Site, West Valley Demonstration Project Site, Los Alamos National Laboratory, Brookhaven National Laboratory, Sandia National Laboratories, General Atomics, San Diego, Babcock ampersand Wilcox, Lynchburg Technical Center, Argonne National Laboratory - East, Naval Reactors Facilities, Rocky Flats Critical Mass Laboratory, EG ampersand G Mound Applied Technologies, Ohio, Lawrence Berkeley Laboratory, and Battelle Columbus Laboratory. This volume also contains information received from the sites that were not visited. These sites include the Naval Reactor Facility at the INEL, EG ampersand G Mound Applied Technologies, The Catholic University of America, Rocky Flats Site, Lawrence Livermore National Laboratory, Stanford Linear Accelerator Laboratory, Energy Technology Engineering Center, and Lawrence Berkeley Laboratory. Information received through the Chicago Operations Office for University Reactors, Massachusetts Institute of Technology, and Battelle Columbus Laboratory is also included. Materials contained in this volume consist of information, data and site documents. They are unedited

  20. Electrical system regulations of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mello, Jose Roberto de; Madi Filho, Tufic, E-mail: jrmello@ipen.br, E-mail: tmfilho@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

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

  1. Scientific Opinion on the safety assessment of the process LPR based on EREMA Advanced and Colortronic SSP ® technology used to recycle post-consumer PET into food contact materials

    Directory of Open Access Journals (Sweden)

    EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids (CEF

    2014-02-01

    Full Text Available This scientific opinion of the EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids deals with the safety assessment of the recycling process LPR (EU register No RECYC061 which is based on the EREMA advanced and Colortronic SSP ® technologies. The input to the process is hot caustic washed and dried PET flakes originating from collected post-consumer PET bottles and containing no more than 5 % of PET from non-food consumer applications. In this process, washed and dried PET flakes are heated successively in two continuous reactors under vacuum before being extruded into pellets. After extrusion they are crystallised and solid state polymerized. Having examined the results of the challenge test provided, the Panel concluded that the four steps, the decontamination in two continuous reactors, extrusion, crystallisation and solid state polymerization are the critical steps that determine the decontamination efficiency of the process. The operating parameters to control the performance of these critical steps are temperature, pressure, gas flow and residence time. Under these conditions, it was demonstrated that the recycling process is able to ensure that the level of migration of potential unknown contaminants into food is below the modelled migration of 0.1 μg/kg food derived from exposure scenario for infants and 0.15 μg/kg food derived from the exposure scenario for toddlers. The Panel concluded that recycled PET obtained from LPR process is not of safety concern when used to manufacture articles intended for food contact materials applications in compliance with the conditions as specified in the conclusion of the opinion.

  2. Comparative Investigation of River Water Quality by OWQI, NSFWQI and Wilcox Indexes (Case study: the Talar River – IRAN

    Directory of Open Access Journals (Sweden)

    Darvishi Gholamreza

    2016-03-01

    Full Text Available Rivers are considered as one of the main resources of water supply for various applications such as agricultural, drinking and industrial purposes. Also, these resources are used as a place for discharge of sewages, industrial wastewater and agricultural drainage. Regarding the fact that each river has a certain capacity for acceptance of pollutants, nowadays qualitative and environmental investigations of these resources are proposed. In this study, qualitative investigation of the Talar river was done according to Oregon Water Quality Index (OWQI, National Sanitation Foundation Water Quality Index (NSFWQI and Wilcox indicators during 2011–2012 years at upstream, midstream and downstream of the river in two periods of wet and dry seasons. According to the results of OWQI, all of the values at 3 stations and both periods are placed at very bad quality category and the water is not acceptable for drinking purposes. According to NSFWQI, the best condition was related to the upstream station at wet season period (58, medium quality and the worst condition was related to the downstream in wet season period (46, very bad quality. Also the results of Wilcox showed that in both periods of wet season and dry season, the water quality is getting better from upstream station to the downstream station, and according to the index classification, the downstream water quality has shown good quality and it is suitable for agriculture.

  3. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  4. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated

  5. Low-dose targeted complement inhibition protects against renal disease and other manifestations of autoimmune disease in MRL/lpr mice.

    Science.gov (United States)

    Atkinson, Carl; Qiao, Fei; Song, Hongbin; Gilkeson, Gary S; Tomlinson, Stephen

    2008-01-15

    Complement appears to play a dual role in the progression of systemic lupus erythematosus, serving a beneficial role in enhancing immune complex clearance, while serving a pathogenic role in inducing local inflammation. To investigate these different roles of complement in a therapeutic setting, MRL/lpr mice were treated with the targeted murine C3 complement inhibitor, CR2-Crry, from 16 to 24 wk of age (after the development of proteinuria). The targeting moiety, CR2, binds to C3 breakdown products deposited at sites of complement activation and has the potential to provide complement inhibition locally without causing systemic inhibition. Administration of CR2-Crry i.v., at a dose of 0.25 mg once a week, was associated with a significant survival benefit, improved kidney function, and a significant reduction in glomerulonephritis and renal vasculitis. The presence of skin lesions and lung bronchiolar and vascular inflammation was also dramatically reduced by CR2-Crry treatment. CR2-Crry treatment also resulted in a significant reduction in autoantibody production, as measured by anti-dsDNA Ab levels, and did not cause an increase in circulating immune complex levels. These effects on autoimmunity and circulating immune complexes represent significant potential advantages over the use of Crry-Ig in MRL/lpr mice, a systemic counterpart of CR2-Crry. CR2-Crry localized preferentially to the kidneys in 16-wk MRL/lpr mice with a kidney-localized half-life of approximately 24 h. Thus, targeted complement inhibition at the C3 level is an effective treatment in murine lupus, even beginning after onset of disease. PMID:18178863

  6. Low-dose targeted complement inhibition protects against renal disease and other manifestations of autoimmune disease in MRL/lpr mice.

    Science.gov (United States)

    Atkinson, Carl; Qiao, Fei; Song, Hongbin; Gilkeson, Gary S; Tomlinson, Stephen

    2008-01-15

    Complement appears to play a dual role in the progression of systemic lupus erythematosus, serving a beneficial role in enhancing immune complex clearance, while serving a pathogenic role in inducing local inflammation. To investigate these different roles of complement in a therapeutic setting, MRL/lpr mice were treated with the targeted murine C3 complement inhibitor, CR2-Crry, from 16 to 24 wk of age (after the development of proteinuria). The targeting moiety, CR2, binds to C3 breakdown products deposited at sites of complement activation and has the potential to provide complement inhibition locally without causing systemic inhibition. Administration of CR2-Crry i.v., at a dose of 0.25 mg once a week, was associated with a significant survival benefit, improved kidney function, and a significant reduction in glomerulonephritis and renal vasculitis. The presence of skin lesions and lung bronchiolar and vascular inflammation was also dramatically reduced by CR2-Crry treatment. CR2-Crry treatment also resulted in a significant reduction in autoantibody production, as measured by anti-dsDNA Ab levels, and did not cause an increase in circulating immune complex levels. These effects on autoimmunity and circulating immune complexes represent significant potential advantages over the use of Crry-Ig in MRL/lpr mice, a systemic counterpart of CR2-Crry. CR2-Crry localized preferentially to the kidneys in 16-wk MRL/lpr mice with a kidney-localized half-life of approximately 24 h. Thus, targeted complement inhibition at the C3 level is an effective treatment in murine lupus, even beginning after onset of disease.

  7. An intrinsic B cell defect is required for the production of autoantibodies in the lpr model of murine systemic autoimmunity

    Energy Technology Data Exchange (ETDEWEB)

    Sobel, E.S.; Katagiri, T.; Katagiri, K.; Morris, S.C.; Cohen, P.L.; Eisenberg, R.A. (Univ. of North Carolina, Chapel Hill (USA))

    1991-06-01

    Mice homozygous for the gene lpr develop marked lymphadenopathy and a spectrum of autoantibodies closely resembling that of human systemic lupus erythematosus. The unusual T cell phenotype of the expanded lymphocyte population and the T-dependence of several antibodies in this strain have suggested that primary T cell abnormalities underlie the autoimmune syndrome. Using double chimeras, we now show that expression of the lpr gene in B cells is absolutely necessary for autoantibody production. Combinations of anti-Thy 1.2 + C' treated bone marrow from congenic strains of C57BL/6 mice, differing only at the immunoglobulin heavy chain (Igh) and lpr loci, were transferred into lethally irradiated B6/lpr mice. Double chimerism was documented by allotype-specific surface IgD and IgM immunofluorescence assay of peripheral blood and by allotype-specific enzyme-linked immunosorbent assay for total IgM in serum. Despite the presence of both +/+ and lpr B cells, IgM and IgG2a anti-chromatin as well as IgM anti-IgG were entirely the products of lpr B cells. Total serum IgG2a and IgG1 were also dominated by the lpr phenotype but not to the same extent. A similar experiment using B6/lpr-Igha recipients confirmed these findings. Additional experiments in which B6/lpr recipients were infused with ratios of donor bone marrow favoring B6.C20 +/+ over B6/lpr showed that even though +/+ B cells were overrepresented, autoantibodies were only of the lpr allotype. In addition, in the presence of lpr B cells, normal B cells showed little response to an exogenous, T cell-dependent antigen. The data thus indicate that lpr B cells manifest an intrinsic abnormality which is essential for autoantibody production in the lpr model.

  8. Interview with Professor Mark Wilcox.

    Science.gov (United States)

    Wilcox, Mark

    2016-08-01

    Mark Wilcox speaks to Georgia Patey, Commissioning Editor: Professor Mark Wilcox is a Consultant Microbiologist and Head of Microbiology at the Leeds Teaching Hospitals (Leeds, UK), the Professor of Medical Microbiology at the University of Leeds (Leeds, UK), and is the Lead on Clostridium difficile and the Head of the UK C. difficile Reference Laboratory for Public Health England (PHE). He was the Director of Infection Prevention (4 years), Infection Control Doctor (8 years) and Clinical Director of Pathology (6 years) at the Leeds Teaching Hospitals. He is Chair of PHE's Rapid Review Panel (reviews utility of infection prevention and control products for National Health Service), Deputy Chair of the UK Department of Health's Antimicrobial Resistance and Healthcare Associated Infection Committee and a member of PHE's HCAI/AR Programme Board. He is a member of UK/European/US working groups on C. difficile infection. He has provided clinical advice as part of the FDA/EMA submissions for the approval of multiple novel antimicrobial agents. He heads a healthcare-associated infection research team at University of Leeds, comprising approximately 30 doctors, scientists and nurses; projects include multiple aspects of C. difficile infection, diagnostics, antimicrobial resistance and the clinical development of new antimicrobial agents. He has authored more than 400 publications, and is the coeditor of Antimicrobial Chemotherapy (5th/6th/7th Editions, 15 December 2007). PMID:27494150

  9. Coal gasification systems engineering and analysis. Appendix G: Commercial design and technology evaluation

    Science.gov (United States)

    1980-01-01

    A technology evaluation of five coal gasifier systems (Koppers-Totzek, Texaco, Babcock and Wilcox, Lurgi and BGC/Lurgi) and procedures and criteria for evaluating competitive commercial coal gasification designs is presented. The technology evaluation is based upon the plant designs and cost estimates developed by the BDM-Mittelhauser team.

  10. 3D Babcock-Leighton Solar Dynamo Models

    Science.gov (United States)

    Miesch, Mark S.; Hazra, Gopal; Karak, Bidya Binay; Teweldebirhan, Kinfe; Upton, Lisa

    2016-05-01

    We present results from the new STABLE (Surface flux Transport and Babcock Leighton) Dynamo Model. STABLE is a 3D Babcock-Leighton/Flux Transport dynamo model in which the source of poloidal field is the explicit emergence, distortion, and dispersal of bipolar magnetic regions (BMRs). In this talk I will discuss initial results with axisymmetric flow fields, focusing on the operation of the model, the general features of the cyclic solutions, and the challenge of achieving supercritical dynamo solutions using only the Babcock-Leighton source term. Then I will present dynamo simulations that include 3D convective flow fields based on the observed velocity power spectrum inferred from photospheric Dopplergrams. I'll use these simulations to assess how the explicit transport and amplification of fields by surface convection influences the operation of the dynamo. I will also discuss the role of surface magnetic fields in regulating the subsurface toroidal flux budget.

  11. CR1/CR2 deficiency alters IgG3 autoantibody production and IgA glomerular deposition in the MRL/lpr model of SLE.

    Science.gov (United States)

    Boackle, Susan A; Culhane, Kristin K; Brown, Jared M; Haas, Mark; Bao, Lihua; Quigg, Richard J; Holers, V Michael

    2004-03-01

    CR1 and CR2 expression is decreased by approximately 50% on B cells of patients with systemic lupus erythematosus (SLE). Expression is also decreased in the MRL/lpr murine model of SLE prior to the development of clinical disease, suggesting that this alteration may play a role in pathogenesis. To determine whether the decrease in receptor levels affects the development of SLE, we analyzed MRL/lpr mice in which CR1/CR2 expression was altered by gene targeting. Mice from each cohort (Cr2+/+, Cr2+/-, and Cr2-/-) were analyzed biweekly for the development of proteinuria and autoantibodies. Kidneys were examined at 12 and 16 weeks for evidence of immune complex deposition and renal disease. Deficiency of CR1/CR2 did not affect survival or development of renal disease as measured by proteinuria. Mice deficient in CR1/CR2 had significantly lower levels of IgG3 rheumatoid factor (RF) and total serum IgG3, suggesting a specific defect in production of IgG3 in response to endogenous autoantigens. Since IgG3 RF has been associated with the development of vasculitis in this model, we examined the mice for alterations in development of this clinical manifestation. Although there was no difference in the development of ear necrosis among the three groups, renal arteritis was not identified in any of the Cr2+/- mice, whereas it was present in 20% of the Cr2+/- and 40% of the Cr2+/+ mice. Finally, significantly higher levels of IgA were seen in the glomeruli of Cr2+/- mice compared to Cr2+/- or Cr2+/+ mice, suggesting that CR1/CR2 are involved in either the regulation of IgA production or the clearance of IgA immune complexes. Together these data support the concept that alterations in CR1/CR2 expression or function affect the regulation of autoantibody production and/or clearance and may have clinical consequences. PMID:15293881

  12. A Solar Dynamo Model Driven by Mean-Field Alpha and Babcock-Leighton Sources: Fluctuations, Grand-Minima-Maxima and Hemispheric Asymmetry in Sunspot Cycles

    CERN Document Server

    Passos, D; Hazra, S; Lopes, I

    2013-01-01

    Extreme solar activity fluctuations and the occurrence of solar grand minima and maxima episodes, are well established, observed features of the solar cycle. Nevertheless, such extreme activity fluctuations and the dynamics of the solar cycle during Maunder minima-like episodes remain ill-understood. We explore the origin of such extreme solar activity fluctuations and the role of dual poloidal field sources, namely the Babcock-Leighton mechanism and the mean-field alpha effect in the dynamics of the solar cycle. We mainly concentrate on entry and recovery from grand minima episodes such as the Maunder minimum and the dynamics of the solar cycle. We use a kinematic solar dynamo model with a novel set-up in which stochastic perturbations force two distinct poloidal field alpha effects. We explore different regimes of operation of these poloidal sources with distinct operating thresholds, to identify the importance of each. The perturbations are implemented independently in both hemispheres which allows one to ...

  13. Development and testing of a diagnostic system for intelligen distributed control at EBR-2

    International Nuclear Information System (INIS)

    A diagnostic system is under development for demonstration of Intelligent Distributed Control at the Experimental Breeder Reactor (EBR--II). In the first phase of the project a diagnostic system is being developed for the EBR-II steam plant based on the DISYS expert systems approach. Current testing uses recorded plant data and data from simulated plant faults. The dynamical simulation of the EBR-II steam plant uses the Babcock and Wilcox (B ampersand W) Modular Modeling System (MMS). At EBR-II the diagnostic system operates in the UNIX workstation and receives live plant data from the plant Data Acquisition System (DAS). Future work will seek implementation of the steam plant diagnostic in a distributed manner using UNIX based computers and Bailey microprocessor-based control system. 10 refs., 6 figs

  14. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

    International Nuclear Information System (INIS)

    performed at 50 reactor facilities. To be published as approved benchmarks the experiments must be evaluated against agreed technical criteria and reviewed by the IRPhE Technical Review Group. A total of 139 of the 143 evaluations are published as approved benchmarks. The remaining four evaluations are published as draft documents only. New to the handbook are benchmark specifications for selected measurements from the Babcock and Wilcox (B and W) Spectral Shift Reactor Lattice Experiment that was performed to study the nuclear properties of rod lattices moderated by D2O-H2O mixtures. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Argentina, Belgium, Brazil, Canada, P.R. of China, Czech Republic, France, Germany, Hungary, Italy, Japan, Republic of Korea, Russian Federation, Serbia, Slovenia, South Africa, Sweden, Switzerland, United Kingdom, and the United States of America

  15. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  16. Obituary: Horace Welcome Babcock, 1912-2003

    Science.gov (United States)

    Vaughan, Arthur Harris

    2003-12-01

    Horace Welcome Babcock died in Santa Barbara, California on 29 August 2003, fifteen days short of his ninety-first birthday. An acclaimed authority on solar and stellar magnetism and the originator of ingenious advances in astronomical instrumentation in his earlier career, he served as Director of Mount Wilson and Palomar (later Hale) Observatories from 1964 until his retirement in 1978. The founding of the Carnegie Institution of Washington's Las Campanas Observatory in Chile was the culmination of his directorship. Horace was born in Pasadena California on 13 September 1912, the only child of Harold Delos Babcock and Mary G. Henderson. His father, an electrical engineer and physicist by training, had been hired by George Ellery Hale to work at the recently founded Mount Wilson Solar Observatory beginning in 1909. Thus Horace spent much of his boyhood on Mount Wilson in the company of astronomers. Horace developed an early interest in astronomy, worked as a volunteer solar observer at Mount Wilson and published his first paper in 1932, with his father. He was fascinated by fine mechanisms and by optical and electrical instruments. After graduating from Caltech with a degree in structural engineering in 1934, he earned his PhD in astronomy at Lick Observatory in 1938. His dissertation provided the first measurement of the rotational velocity curve and a derivation of the mass-to-luminosity ratio for M31; this work is still cited in reviews of the study of ``dark matter." Horace served as a research assistant at Lick Observatory (1938 39) and an Instructor at the University of Chicago's McDonald and Yerkes Observatories (1939--41) under Otto Struve. He undertook radar-related wartime electronics work at the MIT Radiation Laboratory (1941 42) and then worked on aircraft rocket launchers as part of the Caltech Rocket Project (1942 45). This project brought him into contact with Ira S. Bowen, head of the project's Photographic Division. Impressed with his knowledge of

  17. A Numerical Model of Deuterium and Oxygen-18 Diffusion in the Confined Lower Wilcox Aquifer of the Lower Mississippi Valley (USA)

    Science.gov (United States)

    Currens, B. J.; Sawyer, A. H.; Fryar, A. E.; Parris, T. M.; Zhu, J.

    2015-12-01

    Deuterium and oxygen-18 are routinely used with noble gases and radioisotopes (e.g., 2H, 14C, 36Cl) to infer climate during groundwater recharge. However, diffusion of 2H and 18O between a confined aquifer and bounding aquitards could alter total isotope concentrations and the inferred temperature during recharge if groundwater flow is sufficiently slow. Hendry and Schwartz (WRR 24(10), 1988) explained anomalous 2H and 18O enrichment in the Milk River aquifer of Alberta by analytically modeling isotope diffusion between the lower bounding aquitard and the aquifer. Haile (PhD dissertation, U. Kentucky, 2011) inferred the same mechanism to explain 2H and 18O enrichment along a flowpath in the confined Lower Wilcox aquifer of the northern Gulf Coastal Plain in Missouri and Arkansas. Based on the geologic and hydraulic properties of the Lower Wilcox aquifer, a numerical model has been constructed to determine how diffusion may influence 2H and 18O concentrations in regional aquifers with residence times on the order of 104 to 105 years. The model combines solutions for a 1D forward-in-time, finite-difference groundwater flow equation with an explicit-implicit Crank-Nicholson algorithm for advection and diffusion to solve for flow velocity and isotope concentration. Initial results are consistent with the analytical solution of Hendry and Schwartz (1988), indicating diffusion as a means of isotopic enrichment along regional groundwater flowpaths.

  18. Cutting Edge: IL-23 Receptor Deficiency Prevents the Development of Lupus Nephritis in C57BL/6–lpr/lpr Mice

    OpenAIRE

    Kyttaris, Vasileios C.; Zhang, Zheng; Kuchroo, Vijay K.; Oukka, Mohamed; Tsokos, George C.

    2010-01-01

    IL-17–producing T cells infiltrate kidneys of patients with lupus nephritis, and IL-23–treated lymph node cells from lupus-prone mice may transfer disease to Rag1-deficient mice. In this study, we show that IL-23R–deficient lupus-prone C57BL/6–lpr/lpr mice display decreased numbers of CD3+CD4−CD8− cells and IL-17A–producing cells in the lymph nodes and produce less anti-DNA Abs. In addition, clinical and pathology measures of lupus nephritis are abrogated. The presented experiments document t...

  19. FABRICATION PROCESS AND PRODUCT QUALITY IMPROVEMENTS IN ADVANCED GAS REACTOR UCO KERNELS

    Energy Technology Data Exchange (ETDEWEB)

    Charles M Barnes

    2008-09-01

    A major element of the Advanced Gas Reactor (AGR) program is developing fuel fabrication processes to produce high quality uranium-containing kernels, TRISO-coated particles and fuel compacts needed for planned irradiation tests. The goals of the AGR program also include developing the fabrication technology to mass produce this fuel at low cost. Kernels for the first AGR test (“AGR-1) consisted of uranium oxycarbide (UCO) microspheres that werre produced by an internal gelation process followed by high temperature steps tot convert the UO3 + C “green” microspheres to first UO2 + C and then UO2 + UCx. The high temperature steps also densified the kernels. Babcock and Wilcox (B&W) fabricated UCO kernels for the AGR-1 irradiation experiment, which went into the Advance Test Reactor (ATR) at Idaho National Laboratory in December 2006. An evaluation of the kernel process following AGR-1 kernel production led to several recommendations to improve the fabrication process. These recommendations included testing alternative methods of dispersing carbon during broth preparation, evaluating the method of broth mixing, optimizing the broth chemistry, optimizing sintering conditions, and demonstrating fabrication of larger diameter UCO kernels needed for the second AGR irradiation test. Based on these recommendations and requirements, a test program was defined and performed. Certain portions of the test program were performed by Oak Ridge National Laboratory (ORNL), while tests at larger scale were performed by B&W. The tests at B&W have demonstrated improvements in both kernel properties and process operation. Changes in the form of carbon black used and the method of mixing the carbon prior to forming kernels led to improvements in the phase distribution in the sintered kernels, greater consistency in kernel properties, a reduction in forming run time, and simplifications to the forming process. Process parameter variation tests in both forming and sintering steps led

  20. Babcock Redux: An Ammendment of Babcock's Schematic of the Sun's Magnetic Cycle

    CERN Document Server

    Moore, Ronald L; Sterling, Alphonse C

    2016-01-01

    We amend Babcock's original scenario for the global dynamo process that sustains the Sun's 22-year magnetic cycle. The amended scenario fits post-Babcock observed features of the magnetic activity cycle and convection zone, and is based on ideas of Spruit and Roberts (1983) about magnetic flux tubes in the convection zone. A sequence of four schematic cartoons lays out the proposed evolution of the global configuration of the magnetic field above, in, and at the bottom of the convection zone through sunspot Cycle 23 and into Cycle 24. Three key elements of the amended scenario are: (1) as the net following-polarity field from the sunspot-region omega-loop fields of an ongoing sunspot cycle is swept poleward to cancel and replace the opposite-polarity polar-cap field from the previous sunspot cycle, it remains connected to the ongoing sunspot cycle's toroidal source-field band at the bottom of the convection zone; (2) topological pumping by the convection zone's free convection keeps the horizontal extent of t...

  1. Scientific Opinion on the safety assessment of the process LPR based on EREMA Advanced and Colortronic SSP ® technology used to recycle post-consumer PET into food contact materials

    OpenAIRE

    EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids (CEF)

    2014-01-01

    This scientific opinion of the EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids deals with the safety assessment of the recycling process LPR (EU register No RECYC061) which is based on the EREMA advanced and Colortronic SSP ® technologies. The input to the process is hot caustic washed and dried PET flakes originating from collected post-consumer PET bottles and containing no more than 5 % of PET from non-food consumer applications. In this process, washed and d...

  2. Natural occurrence of Nuc in the sera of autoimmune-prone MRL/lpr mice.

    Science.gov (United States)

    Kanai, Y; Miura, K; Uehara, T; Amagai, M; Takeda, O; Tanuma, S; Kurosawa, Y

    1993-10-29

    We previously established a clone of cells termed KML1-7 which produces a soluble factor that boosts anti-DNA antibody production both in vitro and in vivo across the H-2 barrier. By using the purified protein, termed nucleobindin (Nuc), we cloned cDNA and produced recombinant(r) Nuc in E.coli. Although the purified rNuc showed biological activities such as anti-DNA antibody boosting and DNA binding, there was no evidence that Nuc is really associated with autoimmune status in lupus-prone MRL/lpr mice. Here we report that identification of Nuc was successful from the sera of MRL/lpr mice, but not from those of the substrain MRL/n mice, which show no apparent autoimmune syndrome at the same age of MRL/lpr mice, by means of immunochemical as well as N-terminal amino-acid sequencing methods.

  3. CNSS plant concept, capital cost, and multi-unit station economics

    International Nuclear Information System (INIS)

    United Engineers and Constructors (UE and C) and the Babcock and Wilcox Company (B and W) have performed several studies over the last eight years related to small integral pressurized water reactors. These reactors include the 365 MWt (100 MWe) Consolidated Nuclear Steam Generator (CNSG) and the 1200 MWt Consolidated Nuclear Steam System (CNSS). The studies, mostly performed under contract to the Oak Ridge National Laboratory, have led to a 1250 MWt (400 MWe) Consolidated Nuclear Steam System (CNSS) plant concept, with unique design and cost features. This report contains an update of earlier studies of the CNSS reactor and balance-of-plant concept design, capital costs, and multi-unit plant economics incorporating recent design developments, improvements, and post-TMI-2 upgrades. The economic evaluation compares the total system economic impact of a phased, three stage 400 MWe CNSS implementation program, i.e., a three-unit station, to the installation of a single 1200 MWe Pressurized Water Reactor (PWR) into a typical USA utility system

  4. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  5. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    International Nuclear Information System (INIS)

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs

  6. Correction to Wilcox et al. (2016).

    Science.gov (United States)

    2016-05-01

    Reports an error in "How being busy can increase motivation and reduce task completion time" by Keith Wilcox, Juliano Laran, Andrew T. Stephen and Peter P. Zubcsek (Journal of Personality and Social Psychology, 2016[Mar], Vol 110[3], 371-384). In the article, the affiliation of the author Andrew T. Stephen was incorrectly listed in the byline and the author note. The author is affiliated with the University of Oxford. The author note paragraph "Andrew T. Stephen is now at the University of Oxford" should have been omitted. All versions of this article have been corrected. (The following abstract of the original article appeared in record 2016-11945-002.) This research tests the hypothesis that being busy increases motivation and reduces the time it takes to complete tasks for which people miss a deadline. This effect occurs because busy people tend to perceive that they are using their time effectively, which mitigates the sense of failure people have when they miss a task deadline. Studies 1 and 2 show that when people are busy, they are more motivated to complete a task after missing a deadline than those who are not busy, and that the perception that one is using time effectively mediates this effect. Studies 3 and 4 show that this process makes busy people more likely to complete real tasks than people who are not busy. Study 5 uses data from over half a million tasks submitted by thousands of users of a task management software application to show that busy people take less time to complete a task after they miss a deadline for completing it. The findings delineate the conditions under which being busy can mitigate the negative effects of missing a deadline and reduce the time it takes to complete tasks. (PsycINFO Database Record PMID:27176772

  7. Fate of injected CO2 in the Wilcox Group, Louisiana, Gulf Coast Basin: Chemical and isotopic tracers of microbial-brine-rock-CO2 interactions

    Science.gov (United States)

    Shelton, Jenna L.; McIntosh, Jennifer C.; Warwick, Peter D.; Lee Zhi Yi, Amelia

    2016-01-01

    The “2800’ sandstone” of the Olla oil field is an oil and gas-producing reservoir in a coal-bearing interval of the Paleocene–Eocene Wilcox Group in north-central Louisiana, USA. In the 1980s, this producing unit was flooded with CO2 in an enhanced oil recovery (EOR) project, leaving ∼30% of the injected CO2 in the 2800’ sandstone post-injection. This study utilizes isotopic and geochemical tracers from co-produced natural gas, oil and brine to determine the fate of the injected CO2, including the possibility of enhanced microbial conversion of CO2 to CH4 via methanogenesis. Stable carbon isotopes of CO2, CH4 and DIC, together with mol% CO2 show that 4 out of 17 wells sampled in the 2800’ sandstone are still producing injected CO2. The dominant fate of the injected CO2appears to be dissolution in formation fluids and gas-phase trapping. There is some isotopic and geochemical evidence for enhanced microbial methanogenesis in 2 samples; however, the CO2 spread unevenly throughout the reservoir, and thus cannot explain the elevated indicators for methanogenesis observed across the entire field. Vertical migration out of the target 2800’ sandstone reservoir is also apparent in 3 samples located stratigraphically above the target sand. Reservoirs comparable to the 2800’ sandstone, located along a 90-km transect, were also sampled to investigate regional trends in gas composition, brine chemistry and microbial activity. Microbial methane, likely sourced from biodegradation of organic substrates within the formation, was found in all oil fields sampled, while indicators of methanogenesis (e.g. high alkalinity, δ13C-CO2 and δ13C-DIC values) and oxidation of propane were greatest in the Olla Field, likely due to its more ideal environmental conditions (i.e. suitable range of pH, temperature, salinity, sulfate and iron concentrations).

  8. Comparative Investigation of River Water Quality by OWQI, NSFWQI and Wilcox Indexes (Case study: the Talar River – IRAN)

    OpenAIRE

    Darvishi Gholamreza; Kootenaei Farshad Golbabaei; Ramezani Maedeh; Lotfi Eissa; Asgharnia Hosseinali

    2016-01-01

    Rivers are considered as one of the main resources of water supply for various applications such as agricultural, drinking and industrial purposes. Also, these resources are used as a place for discharge of sewages, industrial wastewater and agricultural drainage. Regarding the fact that each river has a certain capacity for acceptance of pollutants, nowadays qualitative and environmental investigations of these resources are proposed. In this study, qualitative investigation of the Talar riv...

  9. Naja naja atra Venom Protects against Manifestations of Systemic Lupus Erythematosus in MRL/lpr Mice

    Directory of Open Access Journals (Sweden)

    Jiali Zhu

    2014-01-01

    Full Text Available Systemic lupus erythematosus (SLE is an autoimmune disease and effective therapy for this pathology is currently unavailable. We previously reported that oral administration of Naja naja atra venom (NNAV had anti-inflammatory and immune regulatory actions. We speculated that NNAV may have therapeutic effects in MRL/lpr SLE mice. Twelve-week-old MRL/lpr mice received oral administration of NNAV (20, 40, and 80 μg/kg or Tripterygium wilfordii polyglycosidium (10 mg/kg daily for 16 weeks. The effects of NNAV on SLE manifestations, including skin erythema, proteinuria, and anxiety-like behaviors, were assessed with visual inspection and Multistix 8 SG strips and open field test, respectively. The pathology of spleen and kidney was examined with H&E staining. The changes in autoimmune antibodies and cytokines were determined with ELISA kits. The results showed that NNAV protected against the manifestation of SLE, including skin erythema and proteinuria. In addition, although no apparent histological change was found in liver and heart in MRL/lpr SLE mice, NNAV reduced the levels of glutamate pyruvate transaminase and creatine kinase. Furthermore, NNAV increased serum C3 and reduced concentrations of circulating globulin, anti-dsDNA antibody, and inflammatory cytokines IL-6 and TNF-α. NNAV also reduced lymphadenopathy and renal injury. These results suggest that NNAV may have therapeutic values in the treatment of SLE by inhibiting autoimmune responses.

  10. Ac-SDKP ameliorates the progression of lupus nephritis in MRL/lpr mice.

    Science.gov (United States)

    Tan, Hechang; Zhao, Jijun; Wang, Shuang; Zhang, Lili; Wang, Hongyue; Huang, Bin; Liang, Yingjie; Yu, Xueqing; Yang, Niansheng

    2012-12-01

    N-acetyl-seryl-aspartyl-lysyl-proline (Ac-SDKP) is an endogenous tetrapeptide which can inhibit the differentiation, migration and activation of macrophages and suppress the proliferation of fibroblast. This study examined the effects of Ac-SDKP on the progression of lupus nephritis (LN). MRL/lpr mice received subcutaneous infusion of Ac-SDKP (1.0 mg kg(-1) d(-1)) or vehicle through implanted osmotic mini-pumps from 12 to 20 weeks until being euthanized. MRL/MpJ mice served as normal controls. The data indicative of renal inflammation and fibrosis were evaluated before and after treatment. Ac-SDKP-treated MRL/lpr mice showed reduced proteinuria and improved renal function compared with vehicle-treated controls. Ac-SDKP-treated mice demonstrated decreased inflammatory infiltrates of T cells and macrophages in the kidneys as compared to vehicle-treated animals. The treatment also inhibited the activation of NF-κB and production of TNF-α. Despite this, immune complex deposition and plasma anti-dsDNA levels were not statistically different between the two groups. In addition, the treatment inhibited renal expression of TGF-β1, α-SMA and fibronectin as well as the phosphorylation of Smad2/3. Ac-SDKP treatment ameliorated LN through exerting anti-inflammatory and anti-fibrotic effects on MRL/lpr mice, providing therapeutic potential for halting the progression of LN.

  11. 8 CFR 245a.11 - Eligibility to adjust to LPR status.

    Science.gov (United States)

    2010-01-01

    ... Legalization Provisions § 245a.11 Eligibility to adjust to LPR status. An eligible alien, as defined in § 245a.10, may adjust status to LPR status under LIFE Legalization if: (a) He or she properly files,...

  12. Replacement steam generators for Calvert Cliffs, Oconee and future replacement design

    International Nuclear Information System (INIS)

    After the completion of steam generators presently being fabricated, a total of forty replacement steam generators will have been built for fourteen reactor units located at ten reactor sites. This represents approximately $1 billion of manufacture excluding installation costs. Replacement steam generator work began with the initiation of the Millstone 2 steam generator replacement program for Northeast Utilities in 1989. Manufacture is presently underway on replacement recirculating steam generators for Calvert Cliffs Units 1 and 2 plants of Constellation Nuclear (OEM Combustion Engineering) and the once-through steam generators for the Oconee 1, 2 and 3 plants of Duke Power (OEM Babcock and Wilcox). These two sites are the first and second respectively to have applied for and received approval for a life extension of 20 years beyond their original operating license. The application and granting of these license extensions reflects a major change in the nuclear industry over the recent past. The attitude to nuclear power has changed from a relatively defensive strategy to a much more optimistic agenda of utility reorganization, purchase of well performing older plants, replacement of aging components, plant refurbishment, and upgrades and applications for license extension. Possible new plants are also being considered. The paper discusses specific features, attributes, performance and operating experience with replacement steam generators (RSGs) both in service and under construction. Industry issues and design features applicable to future replacement steam generators are also reviewed. (author)

  13. SPACE-R Thermionic Space Nuclear Power System: Design and Technology Demonstration Program

    Science.gov (United States)

    1993-05-01

    This semiannual technical progress report summarizes the technical progress and accomplishments for the Thermionic Space Nuclear Power System (TI-SNPS) Design and Technology Demonstration Program of the prime contractor, Space Power Incorporated (SPI), its subcontractors, and supporting national laboratories during the first half of the government fiscal year (GFY) 1993. SPI's subcontractors and supporting national laboratories include: Babcock & Wilcox for the reactor core and externals; Space Systems/Loral for the spacecraft integration; Thermocore for the radiator heat pipes and the heat exchanger; INERTEK of CIS for the TFE, core elements, and nuclear tests; Argonne National Laboratories for nuclear safety, physics, and control verification; and Oak Ridge National laboratories for materials testing. Parametric trade studies are near completion. However, technical input from INERTEK has yet to be provided to determine some of the baseline design configurations. The INERTEK subcontract is expected to be initiated soon. The point design task has been initiated. The thermionic fuel element (TFE) is undergoing several design iterations. The reactor core vessel analysis and design has also been started.

  14. TRAC-PF1/MOD1 calculations and data comparisons for MIST [Multi-Loop Integral System Test] small-break loss-of-coolant accidents with scaled 10 cm2 and 50 cm2 breaks

    International Nuclear Information System (INIS)

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents (SBLOCAs), loss of feedwater and other transients in Babcock and Wilcox (B and W) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 x 4 (2 hot legs and steam generators, 4 cold legs and reactor-coolant pumps) representation of lowered-loop reactor systems of the B and W design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at Stanford Research Institute. The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are underway at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment have been completed for two transients run in the MIST facility. These are the MIST nominal test. Test 3109AA, a scaled 10 cm2 SBLOCA and Test 320201, a scaled 50 cm2 SBLOCA. Only MIST assessment results are presented in this paper

  15. SPACE-R Thermionic Space Nuclear Power System: Design and Technology Demonstration Program. Semiannual technical progress report for period ending March 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    This Semiannual Technical Progress Report summarizes the technical progress and accomplishments for the Thermionic Space Nuclear Power System (TI-SNPS) Design and Technology Demonstration Program of the Prime Contractor, Space Power Incorporated (SPI), its subcontractors and supporting National Laboratories during the first half of the Government Fiscal Year (GFY) 1993. SPI`s subcontractors and supporting National Laboratories include: Babcock & Wilcox for the reactor core and externals; Space Systems/Loral for the spacecraft integration; Thermocore for the radiator heat pipes and the heat exchanger; INERTEK of CIS for the TFE, core elements and nuclear tests; Argonne National Laboratories for nuclear safety, physics and control verification; and Oak Ridge National laboratories for materials testing. Parametric trade studies are near completion. However, technical input from INERTEK has yet to be provided to determine some of the baseline design configurations. The INERTEK subcontract is expected to be initiated soon. The Point Design task has been initiated. The thermionic fuel element (TFE) is undergoing several design iterations. The reactor core vessel analysis and design has also been started.

  16. Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

    Energy Technology Data Exchange (ETDEWEB)

    Soulami, Ayoub [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Paxton, Dean M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-06-15

    Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validation study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.

  17. Repeated 0.5 Gy gamma-ray irradiation attenuates autoimmune disease in MRL-lpr/lpr mice with up-regulation of regulatory T cells

    International Nuclear Information System (INIS)

    Complete text of publication follows. MRL-lpr/lpr mice present a single gene mutation on the Fas (CD95) gene that leads to reduced signaling for apoptosis. With aging, these mice spontaneously develop autoimmune disease and are used as a model of systemic lupus erythematosus. We previously reported attenuation of autoimmune disease in MRL-lpr/lpr mice by repeated γ-ray irradiation (0.5 Gy each time). In this study, we investigated the mechanisms of this attenuation focusing the highly activated CD3+CD4-CD8-B220+ T cells, which are characteristically involved in autoimmune pathology in these mice. We measured the weight of the spleen and the population of CD3+CD4-CD8-B220+ T cells. Splenomegaly and increase in percentage of CD3+CD4-CD8-B220+ T cells, which occur with aging in non-irradiated mice, were suppressed in irradiated mice. To investigate the function of CD3+CD4-CD8-B220+ T cells, we isolated these cells from splenocytes by magnetic cell sorting. Isolated CD3+CD4-CD8-B220+ T cells were more resistant to irradiation-induced cell death than isolated CD4+ T cells. Although high proliferation rate and IL-6 production were observed in isolated CD3+CD4-CD8-B220+ T cells, the proliferation rate and IL-6 production were lower in the cells isolated from the irradiated mice. Moreover, the production of autoantibodies (anti-collagen antibody and anti-single strand DNA antibody) was also lowered by irradiation. These results indicate that activation of CD3+CD4-CD8-B220+ T cells and progression of pathology would be suppressed by repeated 0.5 Gy γ-ray irradiation. To uncover the mechanism of the immune suppression, we analyzed population of regulatory T cells (CD4+CD25+Foxp3+), which suppress activated T cells and excessive autoimmune responses. Intriguingly, significant increase of the percentage of regulatory T cells was observed in irradiated mice. In conclusion, we found that repeated 0.5 Gy γ-ray irradiation suppresses proliferation rate of CD3+CD4-CD8-B220+ T

  18. CD4(+)B220(+)TCRγδ(+) T cells produce IL-17 in lupus-prone MRL/lpr mice.

    Science.gov (United States)

    Qiu, Feng; Li, Tingting; Zhang, Kui; Wan, Jun; Qi, Xiaokun

    2016-09-01

    Systemic lupus erythematosus is an autoimmune disease with comprehensive immune cell disorders. Recent studies suggested that pro-inflammatory cytokine IL-17 plays important role in lupus, leaving the cellular sources and their pathogenic and physiologic characters largely unknown. In the current study, by using lupus-prone MRL/lpr mice, we demonstrated that Th17 response prevails in lupus disease regarding significantly accumulated serum IL-17, increased IL-17-producing splenocytes, and elevated phospho-STAT3 in CD4(+) T cells. Intracellular staining revealed that unusual CD4(+)B220(+) T cells are major IL-17-producing cells, whereas conventional CD4(+)B220(-) T cells are major IFN-γ-producing cells. Subsequent studies showed that CD4(+)B220(+) cells contains both αβ and γδ T cells in the spleen and thymus of MRL/lpr mice. Further study showed that around 60% of γδ T cells in MRL/lpr mice co-express both B220 and CD4 on their surface, and are the major RORγt(+) cells in MRL/lpr mice. Finally, CD4(+)B220(+) T cells alone do not proliferate, but could enhance the proliferation and IFN-γ-production of conventional CD4(+)B220(-) T cells. Our findings suggest the pathogenic role of unusual CD4(+)B220(+) T cells in lupus disease in MRL/lpr mice according to their IL-17-producing ability and stimulatory function for conventional CD4(+)B220(-) T cells. PMID:27235595

  19. Converting the Audience: A Conversation with Agnes Wilcox

    Science.gov (United States)

    Becker, Becky

    2006-01-01

    This article presents a conversation with Agnes Wilcox, Executive Director of Prison Performing Arts in St. Louis, Missouri, about Prison Performing Arts. Although the average person might balk at the notion of interacting with prison inmates, finding it intimidating, worrisome, or self-sacrificial, for Wilcox, Prison Performing Arts is a…

  20. Prevalence of latent eosinophilia among occupational gardeners at Babcock University, Nigeria

    Institute of Scientific and Technical Information of China (English)

    Ayodele Olushola Ilesanmi; Ginnikachi Jennifer Ekwe; Rosemary Isioma Ilesanmi; Damilola Temitope Ogundele; Jacob Kehinde Akintunde; Oluwasogo Adewole Olalubi

    2016-01-01

    Objective: To determine the level of eosinophils present in the blood and sputum samples, presumably as a result of continual occupational exposure to allergens while on duty, as gardeners at Babcock University, Nigeria. Methods: Haemocytometer and Olympus microscope were utilized to estimate eosino-phils population in 44 blood samples and 21 sputum samples respectively. Results: Relationship between the occurrence of eosinophil in blood and the exposure period among Babcock University gardeners had a positive correlation (r = + 0.08, t=4.55, P Conclusions: The nature and the gardening activities are not a risk factor that signifi-cantly affect eosinophil level but duration of exposure to allergens. However, all safety precautionary kits and wears should be enforced and embraced by the concerned occu-pational gardeners so as to avert and subvert its pre-disposing deleterious effect on them.

  1. White Paper on Data Repository Reorganization Proposal for the xLPR Project

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Hilda B [ORNL; Williams, Paul T [ORNL; Bass, Bennett Richard [ORNL

    2012-09-01

    As the xLPR project moves along, it is important to properly manage the knowledge generated by the different groups. We focus specifically on the knowledge and communications written in files, including general documents, source code and executable files. Data generated through the project are different in nature and, for this reason, need to be treated differently. To that end, ORNL put in place a series of tools that facilitate proper storage and management of project data, document and code changes, group collaboration, knowledge transfer, transparency, accountability and auditability. This paper describes the approaches/tools that we recommend for moving the project forward on knowledge management.

  2. A 3D Babcock-Leighton Solar Dynamo Model

    CERN Document Server

    Miesch, Mark S

    2014-01-01

    We present a 3D kinematic solar dynamo model in which poloidal field is generated by the emergence and dispersal of tilted sunspot pairs (more generally Bipolar Magnetic Regions, or BMRs). The axisymmetric component of this model functions similarly to previous 2D Babcock-Leighton (BL) dynamo models that employ a double-ring prescription for poloidal field generation but we generalize this prescription into a 3D flux emergence algorithm that places BMRs on the surface in response to the dynamo-generated toroidal field. In this way, the model can be regarded as a unification of BL dynamo models (2D in radius/latitude) and surface flux transport models (2D in latitude/longitude) into a more self-consistent framework that captures the full 3D structure of the evolving magnetic field. The model reproduces some basic features of the solar cycle including an 11-yr periodicity, equatorward migration of toroidal flux in the deep convection zone, and poleward propagation of poloidal flux at the surface. The poleward-p...

  3. Comparing dynamic responses of recirculating and once-through steam generators for next generation LWRs

    International Nuclear Information System (INIS)

    In this paper two types of steam generators are under consideration for next-generation (pressurized) light water reactors: a recirculating type and a once-through type. The steady-state and dynamic characteristics of these steam generators were compared to facilitate optimization of a particular reactor system design. To compare, the dynamic responses of the two types, as indicated by the feedwater flow, steam generator level, steam flow, steam pressure, steam enthalpy, primary-side pressure and cold-leg temperature, were assessed using Babcock and Wilcox's Modular Modeling System. The once-through steam generator showed a tremendous flexibility to produce superheated steam under diverse conditions (i.e., constant or variable steam throttle pressure and constant or variable average primary temperature) with excellent speed and accuracy in following the load demand. Since the primary and steam sides are closely coupled with the feedwater, the pressurizer should be sized liberally to lessen the sensitivity of the primary response to feedwater upsets and the reliability of the feedwater train should be enhanced. In contrast, the recirculating steam generator must be operated with variable steam throttle pressure and variable primary average temperature, and the speed and accuracy of following the load demand are not as good. While the recirculation provides an effective cushion for the primary and steam sides from feedwater upsets, it also amplifies the level response caused by upsets in steam pressure and feedwater temperature affecting the level controllability and moisture separation performance. The recirculating steam generator should be designed to incorporate features to improve level controllability by constant-inventory control strategy. Also to survive a reactor-coolant pump trip, the design with one reactor-coolant pump per loop should be considered

  4. The effect of aging upon CE and B and W control rod drives

    International Nuclear Information System (INIS)

    The effect of aging upon the Babcock and Wilcox and Combustion Engineering control rod drive systems has been evaluated as part of the US Nuclear Regulatory Commission Nuclear Plant Aging Research program. Operating experience data for the 1980-1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environments, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and engineered safety feature actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

  5. Valve inlet fluid conditions for pressurizer safety and relief valves for B and W 177-FA and 205-FA plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cartin, L.R.; Winks, R.W.; Merchent, J.W.; Brandt, R.T.

    1982-12-01

    The overpressurization transients for the Babcock and Wilcox Company's 177- and 205-FA units are reviewed to determine the range of fluid conditions expected at the inlet of pressurizer safety and relief valves. The final Safety Analysis Report, extended high-pressure injection, and cold overpressurization events are considered. The results of this review, presented in the form of tables and graphs, provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI PWR Safety and Relief Valve Test Program are representative of those expected in their unit(s).

  6. Factory acceptance of the compressor skids at Samifi-Babcock. All pictures show the second stage compressor skid.

    CERN Multimedia

    G. Perinic

    2001-01-01

    Most recent pictures taken during the factory acceptance of the compressor skids at Samifi-Babcock. All pictures show the second stage compressor skid. Picture two was taken during the leak tests and shows all the pockets around flanges and valves.

  7. Green County Nuclear Power Plant. License application

    International Nuclear Information System (INIS)

    The Green County reactor, a PWR to be supplied by Babcock and Wilcox, will be a baseload generating facility planned to provide for mass transit and other public agency electrical needs. The plant is scheduled for completion by 1983 and will have a generating capacity of about 1200 MW(e). (FS)

  8. GPU v. B and W lawsuit review and its effect on TMI-1 (Docket 50-289)

    Energy Technology Data Exchange (ETDEWEB)

    1983-09-01

    This report documents a review by the Nuclear Regulatory Commission (NRC) staff of the General Public Utilities Corporation, et al. v. the Babcock and Wilcox Company, et al. (GPU v. B and W) lawsuit record to assess whether any of the staff's previous conclusions or their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in light of the information contained in the lawsuit record. Details of the lawsuit record are provided in the appendices contained in Volume II of this report.

  9. GPU v. B and W lawsuit review and its effect on TMI-1 (Docket 50-289)

    International Nuclear Information System (INIS)

    This report documents a review by the Nuclear Regulatory Commission (NRC) staff of the General Public Utilities Corporation, et al. v. the Babcock and Wilcox Company, et al. (GPU v. B and W) lawsuit record to assess whether any of the staff's previous conclusions or their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in light of the information contained in the lawsuit record. Details of the lawsuit record are provided in the appendices contained in Volume II of this report

  10. Effects of Jiedu Quyu Ziyin Recipe on the apoptosis and expressions of bcl-2 and bax mRNA of peripheralblood lymphocyte in MRL/lpr mice%解毒祛瘀滋阴药对MRL/lpr小鼠外周血淋巴细胞凋亡及线粒体跨膜电位的影响

    Institute of Scientific and Technical Information of China (English)

    曹灵勇; 谢志军; 王新昌; 温成平; 范永升

    2010-01-01

    Objective To explore the effects of Jiedu Quyu Ziyin Recipe (JQZR) on the apoptosis and expressions of bcl-2 and bax mRNA of peripheral-blood lymphocyte in MRL/lpr mice. Methods 80 MRL/lpr mice were randomly divided into model group,TCM group, Western medicine group and TCM and Western medicine group,20 mice in each group, meanwhile,20 Kunming mice were selected as normal group, then intragastrically administered normal sodium, JQZR apozem, prednisone suspension and JQZR apozem and prednisone suspension, 0. 5ml every time,once daily for 12 weeks respectively. At the end of the 12th week, peripheral-blood lymphocytes of every mice purified by gradient centrifugation were cultivated for 48 hours ,then the apoptosis was detected by flow cytometry. Furthermore,the expressions of bcl-2 and bax mRNA of peripheral-blood lymphocyte were detected by RT-PCR. Results At Oh or 48h,the apoptosis ratios of PBLC in normal group, TCM group,Western medicine group and TCM and Western medicine group are higher than model group and the differences are significant(P 0. 05) ,even if which are significant between TCM group and TCM and Western medicine group or between model group and TCM and Western medicine group( P 0.05),但中药组和模型组与中西药组比较,差异有显著性(P0.05);西药组与中西药组之间差异无显著性(P>0.05).结论 解毒祛瘀滋阴药能增加MRL/lpr小鼠的PBLC凋亡率并下调其线粒体跨膜电位水平.

  11. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  12. A benzenediamine derivate FC-99 attenuates lupus nephritis in MRL/lpr mice via inhibiting myeloid dendritic cell-secreted BAFF.

    Science.gov (United States)

    Ji, Jianjian; Xu, Jingjing; Li, Fanlin; Li, Xiaojing; Gong, Wei; Song, Yuxian; Dou, Huan; Hou, Yayi

    2016-05-01

    Myeloid dendritic cells (DCs) can produce B-cell-activating factor (BAFF) that modulates survival and differentiation of B cells and plays a pivotal role in the pathogenesis of systemic lupus erythematosus (SLE). Toll-like receptor 4 (TLR4) signaling has important functions in the process of BAFF production. Our previous study showed that a benzenediamine derivate FC-99 possesses anti-inflammation activity and directly interacts with interleukin-1 receptor-associated kinase 4 (IRAK4), which was a pivotal molecule in TLR4 signaling. In this study, we demonstrated that FC-99 attenuated lupus nephritis in the MRL/lpr mice. FC-99 also decreased the levels of total immunoglobulin G (IgG), total IgG2a and IgM in sera, as well as the activation of B cells in the spleens of MRL/lpr mice. Moreover, FC-99 inhibited abnormal activation of myeloid DCs in spleens and reduced the levels of BAFF in sera, spleens, and kidneys of MRL/lpr mice. Furthermore, upon TLR4 stimulation with lipopolysaccharide in vitro, FC-99 inhibited IRAK4 phosphorylation, as well as the activation and BAFF production in murine bone marrow-derived DCs. These data indicate that FC-99 attenuates lupus nephritis in MRL/lpr mice via inhibiting DC-secreted BAFF, suggesting that FC-99 may be a potential therapeutic candidate for the treatment of SLE. PMID:27121231

  13. The influence of solar wind on extratropical cyclones – Part 1: Wilcox effect revisited

    Directory of Open Access Journals (Sweden)

    M. Rybanský

    2009-01-01

    Full Text Available A sun-weather correlation, namely the link between solar magnetic sector boundary passage (SBP by the Earth and upper-level tropospheric vorticity area index (VAI, that was found by Wilcox et al. (1974 and shown to be statistically significant by Hines and Halevy (1977 is revisited. A minimum in the VAI one day after SBP followed by an increase a few days later was observed. Using the ECMWF ERA-40 re-analysis dataset for the original period from 1963 to 1973 and extending it to 2002, we have verified what has become known as the "Wilcox effect" for the Northern as well as the Southern Hemisphere winters. The effect persists through years of high and low volcanic aerosol loading except for the Northern Hemisphere at 500 mb, when the VAI minimum is weak during the low aerosol years after 1973, particularly for sector boundaries associated with south-to-north reversals of the interplanetary magnetic field (IMF BZ component. The "disappearance" of the Wilcox effect was found previously by Tinsley et al. (1994 who suggested that enhanced stratospheric volcanic aerosols and changes in air-earth current density are necessary conditions for the effect. The present results indicate that the Wilcox effect does not require high aerosol loading to be detected. The results are corroborated by a correlation with coronal holes where the fast solar wind originates. Ground-based measurements of the green coronal emission line (Fe XIV, 530.3 nm are used in the superposed epoch analysis keyed by the times of sector boundary passage to show a one-to-one correspondence between the mean VAI variations and coronal holes. The VAI is modulated by high-speed solar wind streams with a delay of 1–2 days. The Fourier spectra of VAI time series show peaks at periods similar to those found in the solar corona and solar wind time series. In the modulation of VAI by solar wind the IMF BZ seems to control the phase of the Wilcox effect and the depth of the VAI minimum. The

  14. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  15. Replacement of shutdown cooling system and repair of reactor pressure vessel nozzle welds at NPP Forsmark unit 1 and unit 2

    International Nuclear Information System (INIS)

    The Forsmark Nuclear Power Plant is located about 150 km north of Stockholm. The plant consists of three units with boiling water reactors. Unit 1 and Unit 2 were put into operation in 1981 and 1982, respectively. Both of these units are identical each having a capacity of 970 MW. Unit 3 was completed in 1985 and has a capacity of 1160 MW. In November 1998 Babcock Noell Nuclear was awarded the contract to replace the pipe-work of the two-sectioned Shutdown Cooling System 321 from the nozzles at the reactor pressure vessel to 10 meters outside the containment. Moreover, the inner and outer isolation valves including the penetrations had to be replaced. Finally, the repair of the RPV (reactor pressure vessel) connecting welds of the System 415 (Feed Water) and System 323 (Emergency Cooling) was to be performed. The work was carried out by a Babcock Noell Nuclear team integrating Swedish companies during the outages May/June 2000 in Forsmark 2 and August/September 2000 in Forsmark 1. In the Forsmark Nuclear Power Plant, Units 1 and 2, 19 RPV nozzle connections were improved successfully. All relevant start-up deadlines could be kept. All new tools and manipulators met the stringent project requirements. The mockup qualification of the equipment and the special personnel training performed in advance proved that such challenging work can be managed despite limited preparation time and planned effectively in order to recognize and avoid possible risks. (authors)

  16. Remission of systemic lupus erythematosus disease activity with regulatory cytokine interleukin (IL)-35 in Murphy Roths Large (MRL)/lpr mice.

    Science.gov (United States)

    Cai, Z; Wong, C K; Dong, J; Chu, M; Jiao, D; Kam, N W; Lam, C W K; Tam, L S

    2015-08-01

    The immunological mechanisms mediated by regulatory cytokine interleukin (IL)-35 are unclear in systemic lupus erythematosus (SLE). We investigated the frequency of CD4(+) CD25(+) forkhead box protein 3 (FoxP3)(+) regulatory T (Treg ) and IL-10(+) regulatory B (Breg ) cells and related immunoregulatory mechanisms in a female Murphy Roths Large (MRL)/lpr mouse model of spontaneous lupus-like disease, with or without IL-35 treatment. A remission of histopathology characteristics of lupus flare and nephritis was observed in the MRL/lpr mice upon IL-35 treatment. Accordingly, IL-35 and IL-35 receptor subunits (gp130 and IL-12Rβ2) and cytokines of MRL/lpr and BALB/c mice (normal controls) were measured. The increased anti-inflammatory cytokines and decreased proinflammatory cytokines were possibly associated with the restoration of Treg and Breg frequency in MRL/lpr mice with IL-35 treatment, compared to phosphate-buffered saline (PBS) treatment. mRNA expressions of Treg -related FoxP3, IL-35 subunit (p35 and EBI3) and soluble IL-35 receptor subunit (gp130 and IL12Rβ2) in splenic cells were up-regulated significantly in IL-35-treated mice. Compared with the PBS treatment group, IL-35-treated MRL/lpr mice showed an up-regulation of Treg -related genes and the activation of IL-35-related intracellular Janus kinase/signal transducer and activator of transcription signal pathways, thereby indicating the immunoregulatory role of IL-35 in SLE. These in vivo findings may provide a biochemical basis for further investigation of the regulatory mechanisms of IL-35 for the treatment of autoimmune-mediated inflammation.

  17. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  18. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  19. A three-dimensional Babcock-Leighton solar dynamo model: Initial results with axisymmetric flows

    Science.gov (United States)

    Miesch, Mark S.; Teweldebirhan, Kinfe

    2016-10-01

    The main objective of this paper is to introduce the STABLE (Surface flux Transport And Babcock-LEighton) solar dynamo model. STABLE is a 3D Babcock-Leighton/Flux Transport dynamo model in which the source of poloidal field is the explicit emergence, distortion, and dispersal of bipolar magnetic regions (BMRs). Here we describe the STABLE model in more detail than we have previously and we verify it by reproducing a 2D mean-field benchmark. We also present some representative dynamo simulations, focusing on the special case of kinematic magnetic induction and axisymmetric flow fields. Not all solutions are supercritical; it can be a challenge for the BL mechanism to sustain the dynamo when the turbulent diffusion near the surface is ⩾ 1012 cm2 s-1. However, if BMRs are sufficiently large, deep, and numerous, then sustained, cyclic, dynamo solutions can be found that exhibit solar-like features. Furthermore, we find that the shearing of radial magnetic flux by the surface differential rotation can account for most of the net toroidal flux generation in each hemisphere, as has been recently argued for the Sun by Cameron and Schüssler (2015).

  20. A Three-Dimensional Babcock-Leighton Solar Dynamo Model: Initial Results with Axisymmetric Flows

    CERN Document Server

    Miesch, Mark S

    2015-01-01

    The main objective of this paper is to introduce the STABLE (Surface flux Transport And Babcock-LEighton) solar dynamo model. STABLE is a 3D Babcock-Leighton/Flux Transport dynamo model in which the source of poloidal field is the explicit emergence, distortion, and dispersal of bipolar magnetic regions (BMRs). Here we describe the STABLE model in more detail than we have previously and we verify it by reproducing a 2D mean-field benchmark. We also present some representative dynamo simulations, focusing on the special case of kinematic magnetic induction and axisymmetric flow fields. Not all solutions are supercritical; it can be a challenge for the BL mechanism to sustain the dynamo when the turbulent diffusion near the surface is $\\geq 10^{12}$ cm$^2$ s$^{-1}$. However, if BMRs are sufficiently large, deep, and numerous, then sustained, cyclic, dynamo solutions can be found that exhibit solar-like features. Furthermore, we find that the shearing of radial magnetic flux by the surface differential rotation ...

  1. Environmental Assessment: Geothermal Energy Geopressure Subprogram. Gulf Coast Well Drilling and Testing Activity (Frio, Wilcox, and Tuscaloosa Formations, Texas and Louisiana)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-09-01

    The Department of Energy (DOE) has initiated a program to evaluate the feasibility of developing the geothermal-geopressured energy resources of the Louisiana-Texas Gulf Coast. As part of this effort, DOE is contracting for the drilling of design wells to define the nature and extent of the geopressure resource. At each of several sites, one deep well (4000-6400 m) will be drilled and flow tested. One or more shallow wells will also be drilled to dispose of geopressured brines. Each site will require about 2 ha (5 acres) of land. Construction and initial flow testing will take approximately one year. If initial flow testing is successful, a continuous one-year duration flow test will take place at a rate of up to 6400 m{sup 3} (40,000 bbl) per day. Extensive tests will be conducted on the physical and chemical composition of the fluids, on their temperature and flow rate, on fluid disposal techniques, and on the reliability and performance of equipment. Each project will require a maximum of three years to complete drilling, testing, and site restoration.

  2. The profile of tuberculosis infection at the Babcock University Teaching Hospital

    Directory of Open Access Journals (Sweden)

    Shobowale E.O

    2016-01-01

    Full Text Available Background: Tuberculosis is the leading cause of death from any single pathogen and it has consistently continued to be a major public health challenge globally. Data show that Nigeria ranks tenth among the 22 high tuberculosis burden countries. Aim: This study intends to describe the profile of tuberculosis infections in Babcock University Teaching Hospital. Methods: This was a retrospective cross sectional study of patients presenting to the Tuberculosis Laboratory of Babcock University Teaching Hospital. Results: Patients presenting to BUTH were 2.29 times more likely to have a positive AFB sputum smear result when compared to samples from Primary Health Care Centers – P = 0.05, χ 2 = 3.83, O.R = 2.29, R.R = 1.17, CI = 1.0 – 5.34. Patients presenting to BUTH were more likely to be HIV positive when compared to those from PHC’s p = 0.00, χ 2 = 24.74, df = 2. Conclusion: The burden of tuberculosis is still high in our environment and challenges in its rapid and accurate diagnosis still remain. In order to strengthen tuberculosis control, attention needs to be placed on rapid diagnosis and prompt treatment.

  3. A critical assembly designed to measure neutronic benchmarks in support of the space nuclear thermal propulsion program

    Science.gov (United States)

    Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.; Selcow, Elizabeth C.; Cerbone, Ralph J.

    1993-01-01

    A reactor designed to perform criticality experiments in support of the Space Nuclear Thermal Propulsion program is currently in operation at the Sandia National Laboratories' reactor facility. The reactor is a small, water-moderated system that uses highly enriched uranium particle fuel in a 19-element configuration. Its purpose is to obtain neutronic measurements under a variety of experimental conditions that are subsequently used to benchmark rector-design computer codes. Brookhaven National Laboratory, Babcock & Wilcox, and Sandia National Laboratories participated in determining the reactor's performance requirements, design, follow-on experimentation, and in obtaining the licensing approvals. Brookhaven National Laboratory is primarily responsible for the analytical support, Babcock & Wilcox the hardware design, and Sandia National Laboratories the operational safety. All of the team members participate in determining the experimentation requirements, performance, and data reduction. Initial criticality was achieved in October 1989. An overall description of the reactor is presented along with key design features and safety-related aspects.

  4. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  5. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  6. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  7. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  8. An integrated maintenance strategy for the Babcock 10E Coal Mill

    Energy Technology Data Exchange (ETDEWEB)

    MacIntyre, J. [University of Sunderland (United Kingdom). Centre for Adaptive Systems; Stansfield, D.; Allot, P.; Harris, M. [National Power plc (United Kingdom)

    1998-07-01

    Coal-fired power station around the world have many common features, including similar types of auxiliary plant. One example of such a common area is the coal milling plant. This paper describes how an integrated approach to maintenance of the Babcock 10E Coal Mill has been developed at National Power`s Blyth `B` Station on the North East coast of England. The paper gives details of the types of mechanical problems experienced with the plant, and the various engineering, maintenance, and monitoring strategies which have been integrated into a comprehensive and effective maintenance strategy for this plant. The paper also gives detailed examples of the application of these techniques, the results obtained from them, and goes on to show how this integrated approach has reaped substantial rewards for the Station in terms of availability, reliability and profitability. (author)

  9. The Role of Magnetic Buoyancy in a Babcock-Leighton Type Solar Dynamo

    Indian Academy of Sciences (India)

    Dibyendu Nandy; Arnab Rai Choudhuri

    2000-09-01

    We study the effects of incorporating magnetic buoyancy in a model of the solar dynamo—which draws inspiration from the Babcock-Leighton idea of surface processes generating the poloidal field. We present our main results here.

  10. The MRL/lpr Mouse Strain as a Model for Neuropsychiatric Systemic Lupus Erythematosus

    Directory of Open Access Journals (Sweden)

    Maria Gulinello

    2011-01-01

    Full Text Available To date, CNS disease and neuropsychiatric symptoms of systemic lupus erythematosus (NP-SLE have been understudied compared to end-organ failure and peripheral pathology. In this review, we focus on a specific mouse model of lupus and the ways in which this model reflects some of the most common manifestations and potential mechanisms of human NP-SLE. The mouse MRL lymphoproliferation strain (a.k.a. MRL/lpr spontaneously develops the hallmark serological markers and peripheral pathologies typifying lupus in addition to displaying the cognitive and affective dysfunction characteristic of NP-SLE, which may be among the earliest symptoms of lupus. We suggest that although NP-SLE may share common mechanisms with peripheral organ pathology in lupus, especially in the latter stages of the disease, the immunologically privileged nature of the CNS indicates that early manifestations of particularly mood disorders maybe derived from some unique mechanisms. These include altered cytokine profiles that can activate astrocytes, microglia, and alter neuronal function before dysregulation of the blood-brain barrier and development of clinical autoantibody titres.

  11. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  12. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air

  13. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  14. Results and Analysis of the Infrastructure Request for Information (DE-SOL-0008318)

    Energy Technology Data Exchange (ETDEWEB)

    Heidrich, Brenden John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-07-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) released a request for information (RFI) (DE-SOL-0008318) for “University, National Laboratory, Industry and International Input on Potential Office of Nuclear Energy Infrastructure Investments” on April 13, 2015. DOE-NE solicited information on five specific types of capabilities as well as any others suggested by the community. The RFI proposal period closed on June 19, 2015. From the 26 responses, 34 individual proposals were extracted. Eighteen were associated with a DOE national laboratory, including Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Idaho National Laboratory (INL), Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL). Oak Ridge National Laboratory (ORNL) was referenced in a proposal as a proposed capability location, although the proposal did not originate with ORNL. Five US universities submitted proposals (Massachusetts Institute of Technology, Pennsylvania State University, Rensselaer Polytechnic Institute, University of Houston and the University of Michigan). Three industrial/commercial institutions submitted proposals (AREVA NP, Babcock and Wilcox (B&W) and the Electric Power Research Institute (EPRI)). Eight major themes emerged from the submissions as areas needing additional capability or support for existing capabilities. Two submissions supported multiple areas. The major themes are: Advanced Manufacturing (AM), High Performance Computing (HPC), Ion Irradiation with X-Ray Diagnostics (IIX), Ion Irradiation with TEM Visualization (IIT), Radiochemistry Laboratories (RCL), Test Reactors, Neutron Sources and Critical Facilities (RX) , Sample Preparation and Post-Irradiation Examination (PIE) and Thermal-Hydraulics Test Facilities (THF).

  15. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  16. Coal geology of the Paleocene-Eocene Calvert Bluff Formation (Wilcox Group) and the Eocene Manning Formation (Jackson Group) in east-central Texas; field trip guidebook for the Society for Organic Petrology, Twelfth Annual Meeting, The Woodlands, Texas, August 30, 1995

    Science.gov (United States)

    Warwick, Peter D.; Crowley, Sharon S.

    1995-01-01

    The Jackson and Wilcox Groups of eastern Texas (fig. 1) are the major lignite producing intervals in the Gulf Region. Within these groups, the major lignite-producing formations are the Paleocene-Eocene Calvert Bluff Formation (Wilcox) and the Eocene Manning Formation (Jackson). According to the Keystone Coal Industry Manual (Maclean Hunter Publishing Company, 1994), the Gulf Coast basin produces about 57 million short tons of lignite annually. The state of Texas ranks number 6 in coal production in the United States. Most of the lignite is used for electric power generation in mine-mouth power plant facilities. In recent years, particular interest has been given to lignite quality and the distribution and concentration of about a dozen trace elements that have been identified as potential hazardous air pollutants (HAPs) by the 1990 Clean Air Act Amendments. As pointed out by Oman and Finkelman (1994), Gulf Coast lignite deposits have elevated concentrations of many of the HAPs elements (Be, Cd, Co, Cr, Hg, Mn, Se, U) on a as-received gm/mmBtu basis when compared to other United States coal deposits used for fuel in thermo-electric power plants. Although regulations have not yet been established for acceptable emissions of the HAPs elements during coal burning, considerable research effort has been given to the characterization of these elements in coal feed stocks. The general purpose of the present field trip and of the accompanying collection of papers is to investigate how various aspects of east Texas lignite geology might collectively influence the quality of the lignite fuel. We hope that this collection of papers will help future researchers understand the complex, multifaceted interrelations of coal geology, petrology, palynology and coal quality, and that this introduction to the geology of the lignite deposits of east Texas might serve as a stimulus for new ideas to be applied to other coal basins in the U.S. and abroad.

  17. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  18. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  19. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  20. In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

    International Nuclear Information System (INIS)

    During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

  1. Research reactor education and training

    International Nuclear Information System (INIS)

    CORYS T.E.S.S. and TECHNICATOME present in this document some of the questions that can be rightfully raised concerning education and training of nuclear facilities' staffs. At first, some answers illustrate the tackled generic topics: importance of training, building of a training program, usable tools for training purposes. Afterwards, this paper deals more specifically with research reactors as an actual training tool. The pedagogical advantages they can bring are illustrated through an example consisting in the description of the AZUR facility training capabilities followed by the detailed experiences CORYS T.E.S.S. and TECHNICATOME have both gathered and keeps on gaining using research reactors for training means. The experience shows that this incomparable training material is not necessarily reserved to huge companies or organisations' numerous personnel. It offers enough flexibility to be adapted to the specific needs of a thinner audience. Thus research reactor staffs can also take advantages of this training method. (author)

  2. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235U or 239Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  3. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    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

  4. The effect of salmeterol and salbutamol on mediator release and skin responses in immediate and late phase allergic cutaneous reactions

    DEFF Research Database (Denmark)

    Petersen, Lars Jelstrup; Skov, P S

    1999-01-01

    clinical and biochemical EAR and LPR in human skin. METHODS: Measurement of wheal and flare reactions to allergen, codeine, and histamine, and LPR (induration) to allergen. Assessment of histamine and prostaglandin D2 (PGD2) release by microdialysis technique in EAR, and measurement of mediators in LPR by...

  5. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  6. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  7. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is to

  8. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    Energy Technology Data Exchange (ETDEWEB)

    Tapp, P.A.

    1992-04-01

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms` performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs.

  9. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    Energy Technology Data Exchange (ETDEWEB)

    Tapp, P.A.

    1992-04-01

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms' performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs.

  10. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  11. An example of Ensemble Kalman Filter data assimilation in a Babcock-Leighton solar dynamo model

    Science.gov (United States)

    Dikpati, Mausumi; Anderson, Jeffrey L.

    2016-05-01

    Atmospheric and oceanic prediction models have been greatly advanced over the past 40 years by using modern data assimilation techniques. Application of similar techniques in solar models started about 7 years ago. However, acceptance of such techniques by the solar community has been slow to develop. In order to make accurate predictions of solar activity as well as reconstruction of certain model parameters that cannot be directly measured, it will be essential to implement sophisticated data assimilation techniques as used by atmospheric and oceanic models. We will present here an example of parameter reconstruction, namely the time variation in meridional flow-speed, done by assimilating data into a Babcock-Leighton solar dynamo model in the framework of NCAR's Data Assimilation Research Testbed (NCAR-DART). By performing many 'Observing System Simulation Experiments' (OSSEs) we find that an optimally good reconstruction in time series of meridional circulation can be obtained by using 16 ensemble members and assimilating one magnetic observation with less than 40 percent observational error. However, the RMS error in reconstruction reduces with increase in ensemble size, increase in number of observations and decrease in observational error. We also find that assimilation of magnetic field observations taken from low-to-mid latitudes at the surface compared to any other locations produces the best reconstruction. We will close by showing that assimilation cycle of 15 days is optimal; generally a longer assimilation cycle deteriorates the results, but the Dynamo DART system needs a minimum time to develop the dynamics.

  12. Safety review, assessment and inspection on research reactors, experimental reactors and nuclear heating reactors

    International Nuclear Information System (INIS)

    The NNSA and its regional office step further strengthened the regulation on the safety of in-service research reactors in 1996. A lot of work has been done on the supervision of safe in rectifying the review and assessment of modified items, the review of operational documents, the treatment of accidents, the establishment of the system for operational experience feedback, daily and routine inspection on nuclear safety. The internal management of the operating organization on nuclear safety was further strengthened, nuclear safety culture was further enhanced, the promotion in nuclear safety and the safety situation for in-service research reactors were improved

  13. Process regime variability across growth faults in the Paleogene Lower Wilcox Guadalupe Delta, South Texas Gulf Coast

    Science.gov (United States)

    Olariu, Mariana I.; Ambrose, William A.

    2016-07-01

    The Wilcox Group in Texas is a 3000 m thick unit of clastic sediments deposited along the Gulf of Mexico coast during early Paleogene. This study integrates core facies analysis with subsurface well-log correlation to document the sedimentology and stratigraphy of the Lower Wilcox Guadalupe Delta. Core descriptions indicate a transition from wave- and tidally-influenced to wave-dominated deposition. Upward-coarsening facies successions contain current ripples, organic matter, low trace fossil abundance and low diversity, which suggest deposition in a fluvial prodelta to delta front environment. Heterolithic stratification with lenticular, wavy and flaser bedding indicate tidal influence. Pervasively bioturbated sandy mudstones and muddy sandstones with Cruziana ichnofacies and structureless sandstones with Ophiomorpha record deposition in wave-influenced deltas. Tidal channels truncate delta front deposits and display gradational upward-fining facies successions with basal lags and sandy tabular cross-beds passing into heterolithic tidal flats and biologically homogenized mudstones. Growth faults within the lower Wilcox control expanded thickness of sedimentary units (up to 4 times) on the downdip sides of faults. Increased local accommodation due to fault subsidence favors a stronger wave regime on the outer shelf due to unrestricted fetch and water depth. As the shoreline advances during deltaic progradation, successively more sediment is deposited in the downthrown depocenters and reworked along shore by wave processes, resulting in a thick sedimentary unit characterized by repeated stacking of shoreface sequences. Thick and laterally continuous clean sandstone successions in the downthrown compartments represent attractive hydrocarbon reservoirs. As a consequence of the wave dominance and increased accommodation, thick (tens of meters) sandstone-bodies with increased homogeneity and vertical permeability within the stacked shoreface successions are created.

  14. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  15. Three-dimensional ground-water modeling in depositional systems, Wilcox Group, Oakwood salt dome area, east Texas

    International Nuclear Information System (INIS)

    The data base includes not only measurements of hydraulic head and hydraulic conductivity but also lithofacies maps constructed in a previous study of Wilcox depositional systems. The Carrizo aquifer is a fairly homogeneous sand sheet overlying the much thicker Wilcox Group, a multiple-aquifer system composed primarily of fluvial channel-fill sand bodies distributed among lower permeability interchannel sands and muds. The interconnectedness of the channel-fill sands, which have predictable values of hydraulic conductivity, strongly influences the rate and direction of ground-water flow. Lateral interconnectedness may depend largely on frequency distributions of channel-fill sands (that is, sand percent). Vertical interconnectedness is apparently poor owing to the horizontal stratification of sand and mud. Simulating observed pressure-depth trends by manipulating values of equivalent vertical hydraulic conductivity (K/sub v/') demonstrates that the ratio of vertical to horizontal conductivity (K/sub v/'/K/sub h/') is very low (about 10-3 to 10-4). Locally high values of K/sub v/' could result in locally rapid vertical flow, which could in turn be detected using pressure-depth residence times of 103 to 104 years in channel-fill facies and 105 to 106 years in interchannel facies. Because Oakwood Dome is apparently surrounded by interchannel facies as a result of syndepositional dome growth, the dome may be essentially isolated from circulating Wilcox ground water. A possible exception is where channel-fill facies appear to touch or come close to the northeast flank, coinciding with a brackish-water plume that apparently results from dissolution of salt of cap rock. The northeast orientation of the plume appears to be caused by sand-body distribution and interconnection. 38 references

  16. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  17. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  18. Disturbance analysis and surveillance system scoping and feasibility system. Final report

    International Nuclear Information System (INIS)

    This report summarizes the results of a disturbance analysis and surveillance system (DASS) scoping and feasibility study conducted by The Babcock and Wilcox Company, Burns and Roe, Incorporated, General Physics Corporation, and Duke Power Company for Sandia Laboratories and the US Department of Energy. The report addresses selection of DASS goals and functions, development of a design concept for a DASS based on monitoring the nuclear plant subsystem functions and states against predetermined targets, and creation of engineering procedures for the design and implementation of a DASS. The validity of the procedures is evaluated based on application to a subset of the DASS functions. It is concluded that the DASS design concept is a feasible, systematic, and modular approach to plant disturbance identification

  19. Windows Calorimeter Control (WinCal) system configuration control board (SCCB) operating procedure

    International Nuclear Information System (INIS)

    This document describes the operating procedure for the System Configuration Control Board (SCCB) performed in support of the Windows Calorimeter Control (WinCal) system. This board will consist of representatives from Babcock and Wilcox Hanford Company Babcock and Wilcox Protec, Inc.; and Lockheed Martin Services, Inc. In accordance with agreements for the joint use of the Babcock and Wilcox Hanford Company calorimeters located in the Hanford Site Plutonium Finishing Plant (PFP) Nondestructive Assay Laboratory, concurrence regarding changes to the WinCal system will be obtained from the International Atomic Energy Agency (IAEA). Further, changes to the WinCal software will be communicated to Los Alamos National Laboratory

  20. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  1. The SAPPHIRE and 50 MT projects at BWXT, Lynchburg, VA

    International Nuclear Information System (INIS)

    When the SAPPHIRE project for the down-blending of HEU material of Khazak origin was initiated in 1996 at BWX Technologies (BWXT) formally Babcock and Wilcox in Lynchburg, VA and the Agency was requested to apply its specially designed safeguards measures to the process with a view to provide assurance to the international community that down-blending had actually taken place as stipulated in the USA-Khazak agreement a learning process was initiated from this effort culminating in the current 50 MT downblending process at the same facility with BWXT, the USA Authorities, and the Agency as partners in this technologically advanced enterprise aimed at the downgrading of a substantial quantity of weapons grade material. In the present paper an overview is provided of the road leading to an effective, and mutually agreeable safeguards approach for carrying out verifications in the sensitive environment of a facility devoted to HEU uranium processing. (author)

  2. The SAPPHIRE and 50 MT projects at BWXT, Lynchburg, VA

    International Nuclear Information System (INIS)

    Full text: When the SAPPHIRE project for the down-blending of HEU material of Khazak origin was initiated in 1996 at BWX Technologies (BWXT) formally Babcock and Wilcox in Lynchburg, VA and the Agency was requested to apply its specially designed safeguards measures to the process with a view to provide assurance to the international community that down-blending had actually taken place as stipulated in the USA-Khazak agreement a learning process was initiated from this effort culminating in the current 50 MT downblending process at the same facility with BWXT, the USA Authorities, and the Agency as partners in this technologically advanced enterprise aimed at the downgrading of a substantial quantity of weapons grade material. In the present paper an overview is provided of the road leading to an effective, and mutually agreeable safeguards approach for carrying out verifications in the sensitive environment of a facility devoted to HEU uranium processing. (author)

  3. Steam generator waterlancing at Darlington NGS (system development and field application)

    International Nuclear Information System (INIS)

    From the initial steam generator (SG) inspections at Darlington Nuclear Generating Station (DNGS), the authors know that the sludge accumulations on the secondary side tubesheets have been minimal. DNGS is a fairly new station but the experience at the older Ontario Hydro plants have shown that significant accumulations will happen. A pro-active strategy has been adopted for maintaining SGs that will minimize corrosion product accumulation and the potential for component degradation. During the four year planned Unit maintenance outages, SGs will be inspected and waterlanced using a waterlance system designed and built by Babcock and Wilcox International. This automated state-of-the-art system also allows fully recorded inspections of the tubesheet/first half-lattice supports. Some of the key elements covered include results of the initial field application (May, 1995), system development and design, system qualification, cleaning performance, and lessons learned for future outages

  4. Fabrication and closure development of nuclear waste containers for storage at the Yucca Mountain, Nevada repository

    International Nuclear Information System (INIS)

    US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 1 fig., 7 tabs

  5. Nuclear reactors and disarmament

    International Nuclear Information System (INIS)

    From a brief analysis of the perspectives of nuclear weapons arsenals reduction, a rational use of the energetic potential of the ogives and the authentic destruction of its warlike power is proposed. The fissionable material conversion contained in the nuclear fuel ogives for peaceful uses should be part of the disarmament agreements. This paper pretends to give an approximate idea on the resources re assignation implicancies. (Author)

  6. The effect of aging upon CE and B ampersand W control rod drives

    International Nuclear Information System (INIS)

    The effect of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) Control Rod Drive (CRD) systems has been evaluated as part of the USNRC Nuclear Plant Aging Research (NPAR) program. Operating experience data for the 1980--1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environment, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and Engineered Safety Feature (ESF) actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

  7. Intermediate leak protection/automatic shutdown for B and W helical coil steam generator

    International Nuclear Information System (INIS)

    The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions

  8. Simulation analysis of static and dynamic characteristics of once-through steam generator in concentric annuli tube

    Institute of Scientific and Technical Information of China (English)

    ZHANG Wei; BIAN Xin-qian; XIA Guo-qing

    2006-01-01

    The once-through steam generator (OTSG) in concentric annuli tube is a new type of steam generator which applies double side to transfer heat. The heat flux between the water of centric tube, outside annuli tube and that of annulus channel is assumed to be equal, and then the steam generator's model is built by lumped parameters with moving boundary. In the basis of the built model, static and dynamic characteristics are analyzed.The static characteristics are proved by experiment results in a 19-tube once-through steam generator of Babcock & Wilcox. The characteristics that the lengths of three regions (subcooled region, nucleate boiling region, superheat region) change with power can be explained by theory analysis. The dynamic characteristics accord with the heat and hydraulics and the results of analysis according to the mechanism.

  9. RA nuclear reactor - revitalisation, renewal and applications

    International Nuclear Information System (INIS)

    This book is meant to give professional support in solving the problem of RA reactor, its revitalisation and renewal, as a special help for decision makers. Facts in favor of restarting RA reactor are prevailing. This report is made of six parts. First part includes an overview of basic properties of research reactors in the world and a discussion concerning their future development. RA reactor parameters are analyzed both with low enriched and highly enriched fuel and it has been concluded that the aim of RA reactor renewal should be to obtaining as high as possible thermal neutron flux density. The second part deals with possible applications of RA reactor in fundamental and applied research programs, commercial applications and its role in education and training programs. The third part discusses application of RA reactor as a source of thermal neutrons for fundamental and applied sciences, especially in the condensed matter physics and development of new materials. The role of RA reactor in development of radiation protection systems is emphasised in part four. Some possible commercial applications of Ra reactor are described in part five: isotope production, and their different applications. Part six deals with education and training of staff, with special accent on scientific international cooperation. Basic conclusions of this material meant for decision makers are: restarting RA reactor is the most reasonable and activities related to its revitalisation and renewal should be continued; this program should include solving the problems of education and training of the staff for reactor operation, improvement and different applications; renewal program should include renewal of the experimental devices as a condition of reactor efficient application immediately after its startup

  10. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  11. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  12. Lupus eritematoso sistémico en ratones MRL lpr/lpm y knockouts del receptor de quimioquina CCR2

    OpenAIRE

    Camarasa Lillo, Natalia

    2009-01-01

    INTRODUCCIÓN El lupus eritematoso sistémico es una enfermedad autoinmune cuya principal manifestación y debut de la enfermedad es la glomerulonefritis mediada por complejos inmunes. Los ratones MRL/MpJ-Fas lpr/J (MRL/lpr) llevan una mutación en el gen Fas de la apoptosis que da lugar a una proliferación de linfocitos autoreactivos y son considerados un modelo de ratón que reproduce muy bien la enfermedad lúpica en el humano, con linfadenopatía asociada a proliferación aberrante de células T,...

  13. Laryngopharyngeal Reflux: Diagnosis, Treatment, and Latest Research

    Directory of Open Access Journals (Sweden)

    Campagnolo, Andrea Maria

    2014-01-01

    Full Text Available Introduction Laryngopharyngeal reflux (LPR is a highly prevalent disease and commonly encountered in the otolaryngologist's office. Objective To review the literature on the diagnosis and treatment of LPR. Data Synthesis LPR is associated with symptoms of laryngeal irritation such as throat clearing, coughing, and hoarseness. The main diagnostic methods currently used are laryngoscopy and pH monitoring. The most common laryngoscopic signs are redness and swelling of the throat. However, these findings are not specific of LPR and may be related to other causes or can even be found in healthy individuals. Furthermore, the role of pH monitoring in the diagnosis of LPR is controversial. A therapeutic trial with proton pump inhibitors (PPIs has been suggested to be cost-effective and useful for the diagnosis of LPR. However, the recommendations of PPI therapy for patients with a suspicion of LPR are based on the results of uncontrolled studies, and high placebo response rates suggest a much more complex and multifactorial pathophysiology of LPR than simple acid reflux. Molecular studies have tried to identify biomarkers of reflux such as interleukins, carbonic anhydrase, E-cadherin, and mucin. Conclusion Laryngoscopy and pH monitoring have failed as reliable tests for the diagnosis of LPR. Empirical therapy with PPIs is widely accepted as a diagnostic test and for the treatment of LPR. However, further research is needed to develop a definitive diagnostic test for LPR.

  14. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  15. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  16. The need to address the larger universe of HEU-fueled reactors, including critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    Full text: The RERTR program has focused thus far primarily on ending shipments of HEU fuel to research reactors. This has resulted in giving highest priority to reactors with steady thermal powers of 1 megawatt or more, because they require regular refuelling. Critical facilities and pulsed reactors can also of serious concern, because some of them contain very large amounts of barely-irradiated HEU and plutonium. They could be costly to convert - and conversion to LEU may be impractical for fast-neutron critical assemblies. An assessment should be carried out first, therefore, as to which are still needed. Critical assemblies are required today primarily to benchmark Monte Carlo neutron-transport codes. Perhaps the world nuclear community could share a few instead of each reactor-design institute having its own. There is also a whole universe of HEU-fuelled pressurized-water reactors used to power submarines and other types of nuclear-powered ships. These reactors collectively require much more HEU fuel each year than research reactors. The risk of HEU diversion from their fuel cycles is not zero but it is difficult for outsiders to discuss conversion because of the fuel designs are classified. This makes the conversion of Russia's civilian icebreaker reactors of particular interest because issues of classified fuel design are less problematic and these reactors load annually fuel containing about 400 kg of U-235. Another reason for interest in developing LEU fuel for these reactors is that the KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant. Finally, the research-reactor community is, in any case, faced with developing fuels that can operate at power-reactor-fuel temperatures because there are a few high-powered research reactors that operate in this temperature range. (author)

  17. Safety review and assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    More operational events were occurred at various research reactors in 1995. The NNSA and its regional offices conducted careful investigation and strict regulation. In order to analyze comprehensively the safety situation of inservice research reactors and find same countermeasures the NNSA convened a meeting of the safety regulation on research reactors and a meeting for change experience of the safety regulation on research reactors that were participated in by the operating organizations in 1995. A lot of work has been done in the respects of propagation of regulations on nuclear safety, education of nuclear safety culture, the investigation and treatment of operational events, the reexamine of operation documents, the implementation of rectifying items on nuclear safety, the daily inspection and routine inspection on nuclear safety and the studying on the extending service life of research reactors etc

  18. TRAC PF1/MOD1 calculations and data comparisons for mist feed and bleed and steam generator tube rupture experiments

    Energy Technology Data Exchange (ETDEWEB)

    Siebe, D.A.; Boyack, B.E.; Steiner, J.L.

    1988-01-01

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (BandW) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 /times/ 4 (two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps) representation of lowered-loop reactor system of the BandW design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other integral experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at SRI International (SRI-2). The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for two transients run in the MIST facility. These are MIST Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. Only MIST assessment results are presented in this paper. The TRAC-PF1/MOD1 calculations completed to date for MIST tests are in reasonable agreement with the data from these tests. Reasonable agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. We believe that correct conclusions will be reached if the code is used in similar applications despite minor code/model deficiencies. 7 refs., 5 figs., 2 tabs.

  19. Research reactors and alternative devices for research

    International Nuclear Information System (INIS)

    This report includes papers on research reactors and alternatives to the research reactors - radioisotopic neutron sources, cyclotrons, D-T neutron generators and small accelerators, used for radioisotope production, neutron activation analysis, material science, applied and basic research using neutron beams. A separate abstract was prepared for each of the 7 papers

  20. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

  1. TRIGA reactor owners' seminar. Papers and abstracts

    International Nuclear Information System (INIS)

    The TRIGA Reactor Owners' Conference was planned with the aim of bringing together a group of persons interested in the ownership and operation of TRIGA reactors in the hope that an interchange of viewpoints, information, and experience would prove of mutual benefit

  2. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  3. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  4. Maintenance and material aspects of DREAM reactor

    International Nuclear Information System (INIS)

    A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

  5. ISIS Training Reactor: A Reactor Dedicated to Education and Training for Students and Professionals

    International Nuclear Information System (INIS)

    Conclusion: • INSTN strategy: complete theoretical courses by practical courses on the ISIS research reactor. • Training courses integrated both in Academic degree programs and continuing education. • 27 hours of training courses have been developed focusing on the practical and safety aspects of reactor operation. • The Education and Training activity became the main activity of ISIS reactor: 400 trainees/year; 360 hours/year; 40% in English. • Remote access to the Training courses: Internet Reactor Laboratory under development to be started from 2014 to broadcast training courses from ISIS reactor to guest institutions

  6. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  7. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  8. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  9. Effective utilization and management of research reactors

    International Nuclear Information System (INIS)

    The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

  10. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    Science.gov (United States)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  11. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær;

    2003-01-01

    measurements are reviewed in detail. In the sequel, possible manipulated variables, such as the hydraulic retention time, the organic loading rate, the sludge retention time, temperature, pH and alkalinity are evaluated with respect to the two main reactor types: high-rate and low-rate. Finally, the different......The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...

  12. Nuclear reactor philosophy and criteria

    International Nuclear Information System (INIS)

    Nuclear power plant safety criteria and principles developed in Canada are directed towards minimizing the chance of failure of the fuel and preventing or reducing to an acceptably low level the escape of fission products should fuel failure occur. Safety criteria and practices are set forth in the Reactor Siting Guide, which is based upon the concept of defence in depth. The Guide specifies that design and construction shall follow the best applicable code, standard or practice; the total of all serious process system failures shall not exceed one in three years; special safety systems are to be physically and functionally separate from process systems and each other; and safety systems shall be testable, with unavailability less than 10-3. Doses to the most exposed member of the public due to normal operation, serious process failures, and dual failures are specified. Licensees are also required to consider the effects of extreme conditions due to airplane crashes, explosions, turbine disintegration, pipe burst, and natural disasters. Safety requirements are changing as nuclear power plant designs evolve and in response to social and economic pressures

  13. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  14. Mechanical core coupling and reactors stability

    International Nuclear Information System (INIS)

    Structural parts of nuclear reactors are complex mechanical systems, able to vibrate with a set of proper frequencies when suitably excited. Cyclical variations in the strain state of the materials, including density perturbations, are produced. This periodic changes may affect reactor reactivity. But a variation in reactivity affects reactor thermal power, thus modifying the temperature field of the abovementiones materials. If the variation in temperature fields is fast enough, thermal-mechanical coupling may produce fast variations in strain states, and this, at its turn, modifies the reactivity, and so on. This coupling between mechanical vibrations of the structure and the materials of the core, with power oscillations of the reactor, not only may not be excluded a priori, but it seems that it has been present in some stage of the incidents or accidents that happened during the development of nuclear reactor technology. The purpose of the present communication is: (a) To review and generalize some mathematical models that were proposed in order to describe thermal-mechanical coupling in nuclear reactors. (b) To discuss some conditions in which significant instabilities could arise, including large amplitude power oscillations coupled with mechanical vibrations whose amplitudes are too small to be excluded by conventional criteria of mechanical design. Enough Certain aspects of thr physical safety of nuclear power reactors, that are objected by people that opposes to the renaissance of nucleoelectric generation, are discussed in the framework of the mathematical model proposed in this paper

  15. Reactor accident analysis and evaluation

    International Nuclear Information System (INIS)

    Reactor Management Division of Korea Advanced Energy Research Institute has, so far, adopted, modified and developed quite a number of large programs for nuclear core analysis. During the course of this work, it was found necessary to employ some standard subroutines for handling data, input procedures, core memory management and search files. Many programs share lots of common subroutines and/or functions with other programs. Above all, some of them are in lack of transmittal. During the installation of big codes for CYBER computer, it has drawn our keen attention that many elementary subroutines are heavily machine-dependent and that their conversion is extremely difficult. After having collected and modified the subroutines to fit in different codes, it was finally named KINEP (KAERI Improved Nuclear Environmental Package). KINEP has been proved to be convenient even for smaller programs for general purpose. The KINEP includes about one hundred subroutines to facilitate data handling, operator communications, storage allocation, decimal input, file maintence and scratch I/O. (Author)

  16. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  17. Space reactors - past, present, and future

    International Nuclear Information System (INIS)

    In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

  18. Safety and environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    This paper deals with those problems concerning safety and environmental aspects of the future fusion reactors (e.g. fuel cycle, magnetic failure, after heat disturbances, radioactive waste and magnetic field)

  19. RA reactor operation and maintenance in 1999, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1999 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  20. RA reactor operation and maintenance in 1998, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1998 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  1. MIT research reactor. Power uprate and utilization

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  2. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

  3. Developments and Tendencies in Fission Reactor Concepts

    Science.gov (United States)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  4. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author)

  5. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

    The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given

  6. Gas-cooled reactors and their applications

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review and discuss the current status and recent progress made in the technology and design of gas-cooled reactors and their application for electricity generation, process steam and process heat production. The meeting was attended by more than 200 participants from 25 countries and International Organizations presenting 34 papers. The technical part of the meeting was subdivided into 7 sessions: A. Overview of the Status of Gas-Cooled Reactors and Their Prospects (2 papers); B. Experience with Gas-Cooled Reactors (5 papers); C. Description of Current GCR Plant Designs (10 papers); D. Safety Aspects (4 papers); E. Gas-Cooled Reactor Applications (3 papers); F. Gas-Cooled Reactor Technology (6 papers); G. User's Perspectives on Gas-Cooled Reactors (4 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. A separate abstract was prepared for each of the 34 presentations of this meeting. Refs, figs and tabs

  7. TITAN program and direct cycle fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Yasuyoshi; Yoshizawa, Yoshio; Nitawaki, Takeshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2000-07-01

    In December 1999, the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (TIT) started a new program for the development of advanced nuclear reactors with small and medium size. TITAN is the acronym for the program. A novel concept of a carbon dioxide cooled direct cycle fast reactor with a Rankin cycle has been proposed as the advanced nuclear reactors and evaluated for an alternative option to liquid metal cooled fast reactors (LMFRs). The use of carbon dioxide as coolant eliminates major safety related problems of sodium cooled fast reactors: positive sodium void reactivity, hazardous reaction between sodium and water or air. The decay heat is passively removed by allocating a storage tank of liquidized carbon dioxide between the regenerator and the condenser, and by introducing naturally the carbon dioxide vaporized from the tank into the core in the event of the depressurization accident. The direct cycle results in considerable simplification of the heat transport system owing to the absence of intermediate cooling and water-steam loops comparing with the LMFRs. The thermal efficiency of the direct cycle is evaluated as 34.3 %, which is slightly higher than those in the current BWRs and PWRs. (author)

  8. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    Science.gov (United States)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  9. Paradoxical effects of all-trans-retinoic acid on lupus-like disease in the MRL/lpr mouse model.

    Directory of Open Access Journals (Sweden)

    Xiaofeng Liao

    Full Text Available Roles of all-trans-retinoic acid (tRA, a metabolite of vitamin A (VA, in both tolerogenic and immunogenic responses are documented. However, how tRA affects the development of systemic autoimmunity is poorly understood. Here we demonstrate that tRA have paradoxical effects on the development of autoimmune lupus in the MRL/lpr mouse model. We administered, orally, tRA or VA mixed with 10% of tRA (referred to as VARA to female mice starting from 6 weeks of age. At this age, the mice do not exhibit overt clinical signs of lupus. However, the immunogenic environment preceding disease onset has been established as evidenced by an increase of total IgM/IgG in the plasma and expansion of lymphocytes and dendritic cells in secondary lymphoid organs. After 8 weeks of tRA, but not VARA treatment, significantly higher pathological scores in the skin, brain and lung were observed. These were accompanied by a marked increase in B-cell responses that included autoantibody production and enhanced expression of plasma cell-promoting cytokines. Paradoxically, the number of lymphocytes in the mesenteric lymph node decreased with tRA that led to significantly reduced lymphadenopathy. In addition, tRA differentially affected renal pathology, increasing leukocyte infiltration of renal tubulointerstitium while restoring the size of glomeruli in the kidney cortex. In contrast, minimal induction of inflammation with tRA in the absence of an immunogenic environment in the control mice was observed. Altogether, our results suggest that under a predisposed immunogenic environment in autoimmune lupus, tRA may decrease inflammation in some organs while generating more severe disease in others.

  10. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

  11. The construction, installation and commissioning of the PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    A TRIGA Mark II research reactor has been installed at the Tun Ismail Atomic Research Centre (PUSPATI), Selangor, Malaysia. The reactor was commissioned in July 1982. With the commissioning of the reactor, a new era in the development of nuclear science and technology in Malaysia has just begun. This report describes the construction, installation and commissioning of the reactor. (author)

  12. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  13. TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

    Science.gov (United States)

    Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

    Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

  14. Styrene-maleic anhydride copolymerization in a recycle tubular reactor: reactor stability and product quality

    OpenAIRE

    Belkhiria, Sahbi; Meyer, Thierry; Renken, Albert

    1994-01-01

    A tubular recycle reactor was developed to ensure good homogeneity of concn. and temp. in the copolymn. of styrene and maleic anhydride. The compn. of the copolymer obtained is in good agreements with predicted values and the uniformity of compn. was measured for the entire mol.-wt. distribution. The characterization of the reactor (both hydrodynamic and stability) and the quality of the resulting polymer are presented herein. The limits of use of this reactor for the styrene-maleic anhydride...

  15. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  16. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    Energy Technology Data Exchange (ETDEWEB)

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle

  17. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær;

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...... measurements are reviewed in detail. In the sequel, possible manipulated variables, such as the hydraulic retention time, the organic loading rate, the sludge retention time, temperature, pH and alkalinity are evaluated with respect to the two main reactor types: high-rate and low-rate. Finally, the different...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

  18. Oklo reactors and implications for nuclear science

    CERN Document Server

    Davis, E D; Sharapov, E I

    2014-01-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_...

  19. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

    New options for the development of the nuclear energy economy which might become available by a successful development of fusion-breeders or fusion-fission hybrid power reactors, identified and their nonproliferative attributes are discussed. The more promising proliferation-resistance ettributes identified include: (1) Justification for a significant delay in the initiation of fuel processing, (2) Denaturing the plutonium with 238Pu before its use in power reactors of any kind, and (3) Making practical the development of denatured uranium fuel cycles and, in particular, denaturing the uranium with 232U. Fuel resource utilization, time-table and economic considerations associated with the use of fusion-breeders are also discussed. It is concluded that hybrid reactors may enable developing a nuclear energy economy which is more proliferation resistant than possible otherwise, whileat the same time, assuring high utilization of t he uranium and thorium resources in an economically acceptable way. (author)

  20. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  1. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  2. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  3. Geoneutrinos and reactor antineutrinos at SNO+

    CERN Document Server

    Baldoncini, M; Wipperfurth, S A; Fiorentini, G; Mantovani, F; McDonough, W F; Ricci, B

    2016-01-01

    In the heart of the Creighton Mine near Sudbury (Canada), the SNO+ detector is foreseen to observe almost in equal proportion electron antineutrinos produced by U and Th in the Earth and by nuclear reactors. SNO+ will be the first long baseline experiment to measure a reactor signal dominated by CANDU cores ($\\sim$55\\% of the total reactor signal), which generally burn natural uranium. Approximately 18\\% of the total geoneutrino signal is generated by the U and Th present in the rocks of the Huronian Supergroup-Sudbury Basin: the 60\\% uncertainty on the signal produced by this lithologic unit plays a crucial role on the discrimination power on the mantle signal as well as on the geoneutrino spectral shape reconstruction, which can in principle provide a direct measurement of the Th/U ratio in the Earth.

  4. A coupled $2\\times2$D Babcock-Leighton solar dynamo model. II. Reference dynamo solutions

    CERN Document Server

    Lemerle, Alexandre

    2016-01-01

    In this paper we complete the presentation of a new hybrid $2\\times2$D flux transport dynamo (FTD) model of the solar cycle based on the Babcock-Leighton mechanism of poloidal magnetic field regeneration via the surface decay of bipolar magnetic regions (BMRs). This hybrid model is constructed by allowing the surface flux transport (SFT) simulation described in Lemerle et al. 2015 to provide the poloidal source term to an axisymmetric FTD simulation defined in a meridional plane, which in turn generates the BMRs required by the SFT. A key aspect of this coupling is the definition of an emergence function describing the probability of BMR emergence as a function of the spatial distribution of the internal axisymmetric magnetic field. We use a genetic algorithm to calibrate this function, together with other model parameters, against observed cycle 21 emergence data. We present a reference dynamo solution reproducing many solar cycle characteristics, including good hemispheric coupling, phase relationship betwe...

  5. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [Phenix Plant (France)

    2007-07-01

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  6. Oklo natural reactors: geological and geochemical conditions

    International Nuclear Information System (INIS)

    Published as well as unpublished material on the Oklo natural reactors in Gabon was evaluated with regard to the long-term aspects of nuclear waste disposal. Even though the vast data base available at present can provide only a site specific description of the phenomenon, already this material gives relevant information on plutonium retention, metamictization, fission product release, hydrogeochemical stability and migration of fission products. Generalized conclusions applicable to other nuclear waste repository would require the quantitative reconstruction of t s coupled thermo-hydrologic-chemical processes. This could be achieved by studying the deviations in the 2H/1H and 18O/16O ratios of minerals at Oklo. A further generalization of the findings from Oklo could be realized by examining the newly-discovered reactor zone 10, which was active under very different thermal conditions than the other reactors. 205 refs

  7. Distributed computing and nuclear reactor analysis

    International Nuclear Information System (INIS)

    Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations

  8. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  9. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  10. Comparisons of prediction methods for peak cladding temperature and effective thermal conductivity in spent fuel assemblies of transportation/storage casks

    International Nuclear Information System (INIS)

    Highlights: • Peak cladding temperature (PCT) of spent fuel were evaluated by various methods. • The methods are Wooton–Epstein correlation, two-region model, and CFD. • Temperature difference between two-region and CFD ranges from −0.2 to 9 K. • CFD could be used to calculate PCT because of over-predicting PCT of two-region. • Application using CFD was conducted for spent fuel assembly used in Republic of Korea. - Abstract: When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature. There are three models to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas’s two-region model, Bahney Lotz’s effective thermal conductivity model, and Wooton–Epstein correlation. The peak cladding temperatures of Babcock and Wilcox (B and W) 15 × 15 PWR spent fuel assembly under helium backfill gas were evaluated by using two-dimensional CFD simulation and compared with two models (Wooton–Epstein correlation, two-region model). The peak cladding temperature difference between the two-region model and CFD simulation ranges from −0.2 K to 9 K. Two-region model over-predicts the measured peak cladding temperature that performs in a spent fuel dry storage cask. Therefore the simulation could be used to calculate peak cladding temperature of spent fuel assemblies. Application using CFD simulation was conducted to investigate the peak cladding temperature and effective thermal conductivity of spent fuel assembly used in Korea NPPs: 16 × 16 (CE type) and 17 × 17 (WH type) PWR spent fuel assembly. CFD simulation results are similar to each other, and the difference of temperature drop between the three arrays occurs slightly in all basket wall temperatures. The effective thermal conductivity calculated from the 16 × 16 PWR spent fuel assembly results was more conservative

  11. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  12. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  13. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  14. Development and demonstration of an advanced extended-burnup fuel-assembly design incorporating urania-gadolinia. Second semi-annual progress report, October 1981-March 1982

    Energy Technology Data Exchange (ETDEWEB)

    Newman, L W; Rombough, C T; Thornton, T A

    1982-08-01

    The Babcock and Wilcox Company, Duke Power Company, and the US Department of Energy are participating in an extended-burnup program for pressurized water reactors that will demonstrate an advanced fuel assembly design. This advanced fuel assembly will use a UO/sub 2/-Gd/sub 2/O/sub 3/ burnable-poison fuel mixture along with other state-of-the-art fuel performance and uranium utilization-enhancing design features that include annular pellets, annealed guide tubes, Zircaloy intermediate grids, and removable upper end fittings. Comparisons of the thermal properties of UO/sub 2/-Gd/sub 2/O/sub 3/ specimens containing 2.98, 5.66, and 8.50 wt % Gd/sub 2/O/sub 3/ with UO/sub 2/ specimens showed that thermal conductivity is the only thermal parameter significantly affected by the addition of Gd/sub 2/O/sub 3/. The milling steps used to prepare UO/sub 2/-Gd/sub 2/O/sub 3/ powder result in a powder that is more active than standard UO/sub 2/ powder. As a result, UO/sub 2/-Gd/sub 2/O/sub 3/ fuel has shown more variability than UO/sub 2/ fuel in as-sintered theoretical density and densification behavior. However, a poreforming material, added to the UO/sub 2/-Gd/sub 2/O/sub 3/ powder mixture before sintering, can be used to achieve the desired density. Measured results from critical experiments were compared with predicted data and confirmed the accuracy of the standard two-group diffusion theory model for predicting global and discrete UO/sub 2/-Gd/sub 2/O/sub 3/ effects when cross-section input is appropriately adjusted. The preliminary first two fuel cycles for lead test assemblies of the advanced design were developed. Irradiation of the lead test assemblies is scheduled to begin in 1983 in Duke Power Company's Oconee Unit 1. An intercalibrated movable incore detector system will be used to monitor the performance of the test assemblies during irradiation.

  15. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  16. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

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

    International Nuclear Information System (INIS)

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

  18. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  19. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  20. Fast Reactor and ADS development in China

    International Nuclear Information System (INIS)

    Conclusion: • The Fukushima accident influence China deeply. “The 12th five years plan and 2020 perspective goal of nuclear safety and radioactive pollution prevention” has been approved which means the nuclear may restart in the near future. • A demonstration fast reactor is under design. • More and more research works will be executed on CEFR

  1. Reactor antineutrino fluxes - status and challenges

    CERN Document Server

    Huber, Patrick

    2016-01-01

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  2. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes. PMID:11402837

  3. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

  4. Double Chooz and Reactor Theta13 Experiments

    CERN Document Server

    ,

    2016-01-01

    This is a contribution paper from the Double Chooz experiment to the special issue of NPB on neutrino oscillations. The physics and history of the reactor theta13 experiments, as well as Double Chooz experiment and its neutrino oscillation analyses are reviewed.

  5. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

  6. Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

    International Nuclear Information System (INIS)

    Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

  7. Reactor dosimetry and RPV life management

    Energy Technology Data Exchange (ETDEWEB)

    Belousov, S.; Ilieva, K.; Mitev, M. [Inst. for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Tsarigradsko 72, 1784 Sofia (Bulgaria)

    2011-07-01

    Reactor dosimetry (RD) is a tool that provides data for neutron fluence accumulated over the reactor pressure vessel (RPV) during the reactor operation. This information, however, is not sufficient for RPV lifetime assessment. The life management of RPV is a multidisciplinary task. To assess whether the RPV steel properties at the current stage (for actual accumulated neutron fluence) of reactor operation are still 'safe enough,' the dependence of material properties on the fluence must be known; this is a task for material science (MS). Moreover, the mechanical loading over the RPV during normal operation and accidence have to be known, as well, for evaluation, if the RPV material integrity in this loading condition and existing cracks is provided. The crack loading path in terms of stress intensity factor is carried out by structural analyses (SA). Pressure and temperature distribution over RPV used in these analyses are obtained from a thermal hydraulic (TH) calculation. The conjunction of RD and other disciplines in RPV integrity assessment is analyzed in accordance with the FFP (fitness for purpose) approach. It could help to improve the efficiency in multi-disciplinary tasks solutions. (authors)

  8. Sandwich reactor lattices and Bloch's theorem

    International Nuclear Information System (INIS)

    The study of the neutron flux distribution in repetitive sandwiches of reactor material leads to results analogous to the 1-dimensional form of Bloch's theorem for the electronic structure in crystals. This principle makes it possible to perform analytically accurate homogenisations of sandwich lattices The method has been extended to cover multi group diffusion and transport theory. (author)

  9. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  10. Prometheus Reactor I and C Software Development Methodology, for Action

    International Nuclear Information System (INIS)

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I and C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I and C Software Development Process Manual and Reactor Module Software Development Plan to NR for information

  11. Papers on reactor physics for operators and unit managers

    International Nuclear Information System (INIS)

    The monograph contains papers submitted to the Dukovany nuclear power plant personnel with the aim of improving professional knowledge of reactor operators and unit managers and helping them in their preparation for state examinations. It presents an easy to understand explanation of phenomena unit control room personnel actually encounter. The following topics are covered: radioactivity, nuclear reactions, nuclear fission and the fate of neutrons in the reactor; delayed neutrons; reactor period, reactivity; subcriticality and transition to criticality; heat generation in the reactor; reactivity coefficients; reactivity effects during the fuel cycle; reactivity compensation during power changes; reactor response to reactivity changes; xenon poisoning, samarium poisoning; residual power; unit start-up after refuelling; unit power rise to the minimal controllable level following emergency shutdown; shutdown concentrations; reactor control and safety system; scram rod drop; neutron sensors in the reactor; monitoring system inside the reactor; 3rd unit computer; ''operator's ten commandments''. (P.A.). 36 figs., 2 tabs., 6 refs

  12. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  13. Inherently safe reactors and a second nuclear era.

    Science.gov (United States)

    Weinberg, A M; Spiewak, I

    1984-06-29

    The Swedish PIUS reactor and the German-American small modular high-temperature gas-cooled reactor are inherently safe-that is, their safety relies not upon intervention of humans or of electromechanical devices but on immutable principles of physics and chemistry. A second nuclear era may require commercialization and deployment of such inherently safe reactors, even though existing light-water reactors appear to be as safe as other well-accepted sources of central electricity, particularly hydroelectric dams.

  14. OrthoImagery Submission for Wilcox County, GA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — NAIP imagery is available for distribution within 60 days of the end of a flying season and is intended to provide current information of agricultural conditions in...

  15. DCS Terrain for Wilcox County GA MapMod08

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — Terrain data, as defined in FEMA Guidelines and Specifications, Appendix M: Data Capture Standards, describes the digital topographic data that was used to create...

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

    International Nuclear Information System (INIS)

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

  17. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

    International Nuclear Information System (INIS)

    During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

  18. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  19. RA Reactor operation and maintenance (I-IX), Part I

    International Nuclear Information System (INIS)

    The report on RA reactor operation and maintenance for year 1963 is divided in six tasks. This volume contains the introductory report, and three tasks of the final report, namely reactor exploitation, reactivity changes of the RA reactor before repair, planning of refuelling

  20. Research reactors for the social safety and prosperous neutron use

    International Nuclear Information System (INIS)

    The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

  1. R- AND P- REACTOR BUILDING IN-SITU DECOMISSIONING VISUALIZATION

    Energy Technology Data Exchange (ETDEWEB)

    Bobbitt, J.; Vrettos, N.; Howard, M.

    2010-06-15

    During the early 1950s, five production reactor facilities were built at the Savannah River Site. These facilities were built to produce materials to support the building of the nation's nuclear weapons stockpile in response to the Cold War. R-Reactor and P-Reactor were the first two facilities completed in 1953 and 1954.

  2. A proposed paradigm for solar cycle dynamics mediated via turbulent pumping of magnetic flux in Babcock-Leighton type solar dynamos

    CERN Document Server

    Hazra, Soumitra

    2016-01-01

    At present, Babcock-Leighton flux transport solar dynamo models appear as the most promising model for explaining diverse observational aspects of the sunspot cycle. The success of these flux transport dynamo models is largely dependent upon a single-cell meridional circulation with a deep equatorward component at the base of the Sun's convection zone. However, recent observations suggest that the meridional flow may in fact be very shallow (confined to the top 10 % of the Sun) and more complex than previously thought. Taken together these observations raise serious concerns on the validity of the flux transport paradigm. By accounting for the turbulent pumping of magnetic flux as evidenced in magnetohydrodynamic simulations of solar convection, we demonstrate that flux transport dynamo models can generate solar-like magnetic cycles even if the meridional flow is shallow. Solar-like periodic reversals is recovered even when meridional circulation is altogether absent, however, in this case the solar surface m...

  3. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  4. 400-MWe consolidated nuclear steam system (CNSS). Conceptual design. Executive summary

    International Nuclear Information System (INIS)

    There are a number of small nuclear unit concepts under active study. These include the Process Inherent Ultimate Safety (PIUS) unit and a smaller version of the High-Temperature Gas-Cooled Reactor (HTGR). This study has focused on the Consolidated Nuclear Steam System (CNSS) plant concept. Studies performed by The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) starting in 1974 have led to a 400 MW PWR CNSS plant concept of compact design. Recent economic studies for the CNSS plant show that it offers economic advantages for electric power generation in certain situations. This executive summary presents the results of these studies

  5. Fast reactor technology innovation and visualization

    International Nuclear Information System (INIS)

    Innovations in safety, operations, and maintenance for improving the availability, reliability, and capital cost of the sodium fast reactor are described. Concerning safety these innovations deal with on-line limiting safety settings, inherent core protection, detection of subassembly coolant mis-allocation. Concerning reactor operations these innovations deal with advanced energy conversion, adapting non-base load nuclear plants and on-line diagnostics. Other innovations concern inspection, servicing, refueling. The development of these innovations rely on visualization technology for their use and for demonstration of improvements achievable. A visualization platform for running these innovations and the nuclear plant thermal-hydraulic, structure, and process codes that underlie them are described. The platform hardware consists of a large-scale tiled display and a haptic hand-controller and in the future will grow to include a high-speed network and multiple graphics-client systems

  6. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor

    OpenAIRE

    Chen, Pin-Chuan; Park, Daniel S.; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Steven A Soper; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-01-01

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperatu...

  7. Pakistan research reactor and its utilization

    International Nuclear Information System (INIS)

    The 5 MW enriched uranium fuelled, light water moderated and cooled Pakistan Research reactor became critical on 21st December, 1965 and was taken to full power on 22nd June, 1966. Since then is has been operated for about 23000 hours till 30th June, 1983 without any major break down. It has been used for the studies of neutron cross-sections, nuclear structure, fission physics, structure of material, radiation damage in crystals and semiconductors, studies of geological, biological and environmental samples by neutron activation techniques, radioisotope production, neutron radiography and for training of scientists, engineers and technicians. In the paper we have described briefly the facility of Pakistan Research Reactor and the major work carried around it during the last decade. (author)

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

    International Nuclear Information System (INIS)

    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)

  9. Small and medium power reactors 1985

    International Nuclear Information System (INIS)

    This report is intended for designers and planners concerned with Small and Medium Power Reactors. It provides a record of the presentations during the meetings held on this subject at the Agency's General Conference in September 1985. This information should be useful as it indicates the principal findings and main conclusions and recommendations resulting from these meetings. A separate abstract was prepared for each of the 10 presentations in this report

  10. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

  11. Selection and challenges for LFR reactor materials

    International Nuclear Information System (INIS)

    Nuclear energy using Fast GenIV reactors can fulfil future demands concerning CO2 free, base load capability and sustainability. One of the most promising coolants especially due to its high thermal inertia is liquid lead (Pb). Since several years researches all over the world investigate this coolant and its impact on the reactor design and by that on the materials to be selected. The LEADER project, a follow up of ELSY, aims to design a prototypical demonstrator ALFRED and to continue with several design related aspects of the ELFR reactor. For a demonstrator the criteria of material selection are somewhat different to a commercial type like the ELFR. Material selection for ELFR of course considers all the aspects relevant for ALFRED plus the targeted burn up and the expected total dpa related damage especially of the fuel pins. In the past compatibility of structural material (steels like 316L, T91 and 15-15Ti (1.4970)) that can be employed for Pb cooled fast nuclear reactors were investigated in several EU projects like EUROTRANS and worldwide. Solubility of steel alloying elements like Ni, Fe, Cr is the driving force for the reduced corrosion resistance in contact with Pb. In-situ oxidation is the acknowledged measure to protect steels in Pb up to certain temperatures that are material dependent. Based on experiments and the derived temperature limits the average core outlet temperatures of ALFRED and the ELFR are set to 480 C. The most challenging conditions with respect to temperature are at the fuel assembly and the heat exchangers. For both, thin stable oxide scales with negligible reduction in heat transfer are the requested protection method. This presentation will give an overview on the selected materials for ALFRED and ELFR considering, beside pure compatibility, the influence of mechanical interaction like creep and fretting. (orig.)

  12. The Global Outlook for Small Reactors: Opportunities, Challenges and Implementation

    International Nuclear Information System (INIS)

    The fascinating topic of small nuclear is becoming more prevalent on the nuclear agenda. The discussions are generally focused within the country of technical origin. In this presentation 'The global outlook for small reactors' Rolls-Royce along with energy business analysts Douglas-Westwood present their shared views on the global opportunities for Small Reactor deployment in the context of the wider energy market. The presentation will: provide a compressive overview of trends and dynamics relating to Small Reactors in the context of the current world energy market, identify specific Small Reactor opportunities and areas of interest, address the challenges and potential solutions for Small Reactor deployment and operation.(author).

  13. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  14. Oklo reactors and implications for nuclear science

    Science.gov (United States)

    Davis, E. D.; Gould, C. R.; Sharapov, E. I.

    2014-04-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross-sections are input to all Oklo modeling and we discuss a parameter, the 175Lu ground state cross-section for thermal neutron capture leading to the isomer 176mLu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant α and the ratio Xq = mq/Λ (where mq is the average of the u and d current quark masses and Λ is the mass scale of quantum chromodynamics (QCD)). We suggest a formula for the combined sensitivity to α and Xq that exhibits the dependence on proton number Z and mass number A, potentially allowing quantum electrodynamic (QED) and QCD effects to be disentangled if a broader range of isotopic abundance data becomes available.

  15. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  16. Development of Education and Training Programs Using ISIS Research Reactor

    International Nuclear Information System (INIS)

    As a part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) carries out various education and training programs on nuclear reactor theory and operation. These programs take advantage of the use of an extensive range of training tools that includes software applications, simulators, as well as the use of research reactors. After a presentation of ISIS reactor, we present the training courses that have been developed on ISIS reactor and their use in education and training programs developed by INSTN. We report on how the training courses carried out on ISIS research reactor ensure a practical and comprehensive understanding of the reactor principle and operation, bringing tremendous benefit to the trainees. We also discuss the future development of education and training programs using the ISIS research reactor as a very powerful tool for the development of the human resources needed by the nuclear industry and the nuclear programs. (author)

  17. The present status and the prospect of China research reactors

    International Nuclear Information System (INIS)

    A total of 100 reactor operation years' experience of research reactors has now been obtained in China. The type and principal parameters of China research reactors and their operating status are briefly introduced in this paper. Chinese research reactors have been playing an important role in nuclear power and nuclear weapon development, industrial and agricultural production, medicine, basic and applied science research and environmental protection, etc. The utilization scale, benefits and achievements will be given. There is a good safety record in the operation of these reactors. A general safety review is discussed. The important incidents and accidents happening during a hundred reactor operating years are described and analyzed. China has the capability of developing any type of research reactor. The prospective projects are briefly introduced

  18. Nuclear reactor dismantling method and device

    International Nuclear Information System (INIS)

    The reactor dismantling device according to the present invention comprises an elevator lift extending from the lower portion in a biological-shielding walls of the reactor to an operation floor thereabove and a scaffolding for cutting operation vertically disposed at the periphery thereof. The scaffolding is rotated around a cutting mast by remote control to displace a cutting position for a drilling device. Then, pieces of the biological-shielding walls are cut out and automatically transported from the inside of the biological-shielding walls to an operation chamber by a recovery device, a truck and an elevator lift. This makes the dismantling operation highly efficient, to shorten the term of works. Further, operators' exposure dose can be mitigated, thereby enabling to improve safety of the dismantling operation. (T.M.)

  19. Reactor Simulation for Antineutrino Experiments using DRAGON and MURE

    CERN Document Server

    Jones, C L; Conrad, J M; Djurcic, Z; Fallot, M; Giot, L; Keefer, G; Onillon, A; Winslow, L

    2011-01-01

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulations to predict reactor fission rates. Here we present results from the DRAGON and MURE simulation codes and compare them to other industry standards for reactor core modeling. We use published data from the Takahama-3 reactor to evaluate the quality of these simulations against the independently measured fuel isotopic composition. The propagation of the uncertainty in the reactor operating parameters to the resulting antineutrino flux predictions is also discussed.

  20. Shielding design for research and education reactor

    International Nuclear Information System (INIS)

    For the purpose of education and research at the University, 20-KW powered SLOWPOKE-2 research reactor has been chosen as a prototype reactor. In order to study the safety characteristics of the reactor, exposure rate has been estimated at the pool boundary. Reactor core as a radiation source is assumed to be cylindrical volume source. Thus point kernel integration method can be applied to determine the exposure rate. For the sake of simplicity, calculation was done only for the prompt fission gamma rays and fission product gamma rays. As a result, the maximum exposure rate at the pool boundary was estimated to be 18R/min at the same height of the center of the core. In order to examine the accuracy for the point kernel integration method, two shielding experiments were carried out: one for the water tank only and the other for with concrete blocks outside the water tank. Water tank was made of wood pieces which is 13.4cm wide, 1.5cm thick and 2.15m long. Thus the water tank has the total dimension of 1 m radius and 2.1 m height. The experiment was carried out for the radiation source of 0.968 mCi Co-60 at the center of the water tank and the penetrated gamma rays were measured at 5 different detector positions. For the measurement and analysis of the responses, NaI(T1) 3''x3'' detector and 256 channel multichannel analyzer was utilized. To convert pulse height distribution to the exposure rate, Moriuchi conversion factor was adopted. Data from the calculations by point kernel method were well agreed within 10% band with the data from the the experiments. (Author)

  1. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  2. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 103Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  3. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  4. RA reactor operation and maintenance in 2000, Part 1

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis

  5. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  6. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 1013Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 1010Bq (0.5 Ci) per day per ton of fuel

  7. High temperature reactors and their use in the FRG

    International Nuclear Information System (INIS)

    Various aspects of the strategy of building high temperature reactors in the FRG are discussed. The development of these reactors has a long tradition in the FRG and great sums of money are being invested in the research programme. In 1988 the AVR-15 experimental reactor is expected to be shut down in which the helium output temperature had been maintained at 950 degC for a long period of time. The THTR-300 demonstration power plant which is expected to be available at that time represents a link to further application of high temperature reactors in the FRG. A detailed description is presented of projects of further high temperature reactors with a wide range of power output. The BBC/HRB association with Swiss participation is now specifying the project of the HTR-500 reactor with a steam cycle and the delivery of technological steam. This reactor should be followed up by the construction of a reactor with an HHT gas turbine and of an HTR-PNP reactor for coal gasification. Alternatively developed are small HTR-100 universal reactors. Prospective projects also include the 80 MW modular system by KWU following up on the AVR-15 reactor. (Z.M.)

  8. INVAP Experience in the Design and Construction of Research Reactors. (Research Reactors in and from Argentina)

    International Nuclear Information System (INIS)

    Full Text: Argentina has a long tradition in the design and construction of Research Reactors. The first research reactor in Argentina, RA-1, was built by CNEA (Argentina Atomic Energy Commission) in 1958, using drawings lent by USA. RA-2, RA-3, RA-4 and RA-0 followed through. In 1976, a career degree in Nuclear Engineering was started by CNEA and the University of Cuyo in Bariloche. It was decided that there would be a university type reactor to assist with the training of the students. INVAP, a recently created company, was assigned the task of building the reactor in accordance with the engineering developed by CNEA. The RA-6 was a very successful project, which allowed INVAP to build the knowledge for participating in RR projects abroad. Since 1982, INVAP has built research reactors in Algeria, Egypt, Argentina and Australia and had a large participation in the RRs CNEA built in Peru. INVAP has also designed several other RR for different clients, which were not subsequently built. This paper explores this history, giving details of the RR projects in which INVAP has been involved through the years. (author)

  9. Animal Guts as Ideal Reactors: An Open-Ended Project for a Course in Kinetics and Reactor Design.

    Science.gov (United States)

    Carlson, Eric D.; Gast, Alice P.

    1998-01-01

    Presents an open-ended project tailored for a senior kinetics and reactor design course in which basic reactor design equations are used to model the digestive systems of several animals. Describes the assignment as well as the results. (DDR)

  10. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  11. Regulation concerning installation and operation of reactors for power generation

    International Nuclear Information System (INIS)

    The regulation is defined under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and provisions concerning installation and operation of reactors for power generation in the order for execution of the law. Basic concepts and terms are explained, such as: radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area and employee. The application for permission of installation of reactors shall include location and general structure of reactor facilities, structure and equipment of reactors, handling and storing facilities of nuclear fuel materials and facilities for measurement and control, etc. Operation program of reactors shall be prepared for each reactor according to the form attached and filed every fiscal year from that one when the operation is expected to begin. Records shall be made for each reactor and kept for particular periods on inspection of reactor facilities, operation, fuel assembly, control of radiation and maintenance, etc. Entrance to the controlled area shall be limited through specified measures. Exposure dose, inspection, check up, independent examination and operation of reactors, transport and disposal in the works or the enterprise and others are in detail stipulated. Reports shall be submitted to the Minister of International Trade and Industry on concentration of radioactive materials, exposure dose of employees and other designated matters. (Okada, K.)

  12. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  13. Reviewing reactor engineering and fuel handling

    International Nuclear Information System (INIS)

    Experience has shown that the better operating nuclear power plants have well defined and effectively administered policies and procedures for governing reactor engineering and fuel handling (RE and FH) activities. This document provides supplementary guidance to OSART experts for evaluating the RE and FH programmes and activities at a nuclear power plant and assessing their effectiveness and adequacy. It is in no way intended to conflict with existing regulations and rules, but rather to exemplify those characteristics and features that are desirable for an effective, well structured RE and FH programme. This supplementary guidance addresses those aspects of RE and FH activities that are required in order to ensure optimum core operation for a nuclear reactor without compromising the limits imposed by the design, safety considerations of the nuclear fuel. In the context of this document, reactor engineering refers to those activities associated with in-core fuel and reactivity management, whereas fuel handling refers to the movement, storage, control and accountability of unirradiated and irradiated fuel. The document comprises five main sections and several appendices. In Section 2 of this guide, the essential aspects of an effective RE and FH programme are discussed. In Section 3, the various types of documents and reference materials needed for the preparatory work and investigation are listed. In Section 4, specific guidelines for investigation of RE and FH programmes are presented. In Section 5, the essential attributes of an excellent RE and FH programme are listed. The supplementary guidance is concluded with a series of appendices exemplifying the various qualities and attributes of a sound, well defined RE and FH programme

  14. Reactor decommissioning experience and perspectives

    International Nuclear Information System (INIS)

    This paper first describes the existing market context and available techniques, then reviews the contribution of past and present operations and research before discussing the future orientations necessary to develop the means (cutting tools, decontamination processes, telemanipulation and waste conditioning) required to improve the cost effectiveness of decommissioning nuclear power plants. (author)

  15. Mesh generation technology for nuclear reactor simulation; barriers and opportunities

    International Nuclear Information System (INIS)

    Mesh generation in support of nuclear reactor simulation has much in common with the requirements of other application areas, such as computational fluid dynamics (CFD). Indeed, fluid dynamics analysis of the coolant behavior inside the reactor core is an internal flow problem that requires the resolution of spatial and temporal variations in the flow caused by complex component configurations, fluids/structure interaction, turbulence, and thermal heating of the coolant. Typical concerns of meshing complex geometries; the use of hexahedral vs. tetrahedral elements, element geometric quality, mesh smoothness, use of anisotropic elements in the thermal boundary layer, etc., are all considerations important to the reactor meshing problem. Reactor meshing begins to become more specialized as the need to employ reactor simulation as a predictive design and safety analysis capability grows in importance. First, a predictive capability will require more precise physical models to be included, and these models will need to be supported by a computational science framework that will allow them to be accurately approximated both spatially and temporally during the reactor core analysis. Both the multiphysical nature of the composite reactor model and details of the physics algorithms themselves will place new requirements on the meshing process needed to support multidimensional reactor simulation. This article discusses the current state of meshing technology applied to reactor simulation and examines a set of issues that are important in the generation of high-quality reactor meshes today and in the future

  16. Which future for 3. and 4. generation reactors?

    International Nuclear Information System (INIS)

    After having briefly recalled some characteristics of energy producing nuclear reactors by presenting their three main components (fuel, heat transfer fluid, moderator), and outlined that about twenty types of reactors have been historically tested as prototypes in the USA, Russia, UK and France, the author addresses third generation reactors. He states that these reactors do not display an important technological break with respect to PWRs which are presently exploited in France, but that technical advances are such that one can say they belong to a new generation. He states that the EPR (European pressurized reactor) is amongst the best reactors presently on the market. He outlines its technological advances: safety, increased containment, performance, adaptation to various fuel types, availability, reduction of workers exposure, easier maintenance). Of course, the author evokes construction delays and costs for the Finnish and French reactors. Then, he addresses fourth-generation reactors which comprise six types of system: supercritical water reactors, very high temperature reactors (for non electricity generation applications), and four fast neutron systems. These systems have already been experimented in the past and some will be operated in India and Russia. However, due to the relatively low price of uranium and to the high level of uranium reserves, these fast breeders are not really needed on the short or on the medium term. The author outlines France's commitment in the field of fast breeders

  17. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  18. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  19. EBT concept: status, plans, and reactor potential

    International Nuclear Information System (INIS)

    This paper will consist of four parts. First, I will describe the basic EBT machine and its operation. Second, I will discuss the experimental results obtained on the present machine and how they relate to neoclassical transport theory. This will include a discussion of the chronology of the estimates of the confinement time and why they have changed. Third, I will briefly discuss some preliminary proposals on how to improve the confinement and operation of EBTs, and fourth, I will review some of the potential advantages of EBTs as fusion reactors and the calculations indicating the critical issues

  20. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.

    1994-02-25

    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  1. BDDR, a new CEA technological and operating reactor database

    Energy Technology Data Exchange (ETDEWEB)

    Soldevilla, M.; Salmons, S.; Espinosa, B. [CEA-Saclay, CEA/DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette (France); Clanet, M.; Boudin, X. [CEA-Bruyeres-le-Chatel, 91297 Arpajon (France)

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  2. Reactor numerical simulation and hydraulic test research

    Energy Technology Data Exchange (ETDEWEB)

    Yang, L. S. [Nuclear Power Institute of China, Beijing (China)

    2009-07-01

    In recent years, the computer hardware was improved on the numerical simulation on flow field in the reactor. In our laboratory, we usually use the Pro/e or UG commercial software. After completed topology geometry, ICEM-CFD is used to get mesh for computation. Exact geometrical similarity is maintained between the main flow paths of the model and the prototype, with the exception of the core simulation design of the fuel assemblies. The drive line system is composed of drive mechanism, guide bush assembly, fuel assembly and control rod assembly, and fitted with the rod level indicator and drive mechanism power device.

  3. Media and Australia's replacement reactor project

    International Nuclear Information System (INIS)

    In September 1997, the Commonwealth Government of Australia announced a proposal to build a replacement nuclear research reactor at Lucas Heights in Sydney. Extensive public consultation, parliamentary debate and independent reports were prepared to ensure that the new facility would meet strict international requirements, national safety and environmental standards, and performance specifications servicing the needs of Australia - for decades to come. On 6 June 2000, Argentine company INVAP SE was announced as the preferred tenderer. In July 2000 contracts were signed between INVAP and the Australian Nuclear Science and Technology Organisation for the construction the replacement reactor, due to be completed in 2005. In order to retain a strong local presence, INVAP undertook a joint venture with two of Australia's foremost heavy construction businesses. Briefly the new research reactor will be a replacement for the ageing Australian Reactor (HIFAR). Nuclear science and technology, in Australia, is no stranger to media controversy and misinformation. Understandably the announcement of a preferred tenderer followed by the signing of contracts, attracted significant national and international media attention. However in the minds of the media, the issue is far from resolved and is now a constant 'news story' in the Australian media. Baseless media stories have made claims that the project will cost double the original estimates; question the credibility of the contractors; and raise issues of international security. The project is currently linked with Australia's requirements for long term nuclear waste management and there has been an attempt to bring national Indigenous People's issues into play. Some of these issues have been profiled in the press internationally. So, just to set the record straight and give you an appropriate impression of what's 'really happening' I would like to highlight a few issues, how ANSTO dealt with these, and what was finally reported

  4. International Reactor Innovative and Secure (IRIS) summary

    International Nuclear Information System (INIS)

    The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

  5. Small fusion reactors: problems, promise, and pathways

    International Nuclear Information System (INIS)

    The prevalent vision of magnetic fusion as a central-station power plant projects a high-technology, low-power-density nuclear boiler that may require high energy costs to be economic. Smaller, higher-power-density approaches can reduce the impact of the fusion power core and associated support equipment on the overall cost equation for fusion. In the course of attaining sizes, power capacity, and costs that are more in line with alternative energy sources, a range of problems, promise, and pathways can be identified. The issues related to these more compact systems are addressed on the basis of generic reactor models

  6. Research about reactor operator's personality characteristics and performance

    International Nuclear Information System (INIS)

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  7. Anaerobic granular sludge and biofilm reactors

    DEFF Research Database (Denmark)

    Skiadas, Ioannis V.; Gavala, Hariklia N.; Schmidt, Jens Ejbye;

    2003-01-01

    by the immobilization of the biomass, which forms static biofilms, particle-supported biofilms, or granules depending on the reactor's operational conditions. The advantages of the high-rate anaerobic digestion over the conventional aerobic wastewater treatment methods has created a clear trend for the change......-rate anaerobic treatment systems based on anaerobic granular sludge and biofilm are described in this chapter. Emphasis is given to a) the Up-flow Anaerobic Sludge Blanket (UASB) systems, b) the main characteristics of the anaerobic granular sludge, and c) the factors that control the granulation process...

  8. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  9. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

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

    International Nuclear Information System (INIS)

    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

  11. Dynamic reactor modeling with applications to SPR and ZEDNA.

    Energy Technology Data Exchange (ETDEWEB)

    Suo-Anttila, Ahti Jorma

    2011-12-01

    A dynamic reactor model has been developed for pulse-type reactor applications. The model predicts reactor power, axial and radial fuel expansion, prompt and delayed neutron population, and prompt and delayed gamma population. All model predictions are made as a function of time. The model includes the reactivity effect of fuel expansion on a dynamic timescale as a feedback mechanism for reactor power. All inputs to the model are calculated from first principles, either directly by solving systems of equations, or indirectly from Monte Carlo N-Particle Transport Code (MCNP) derived results. The model does not include any empirical parameters that can be adjusted to match experimental data. Comparisons of model predictions to actual Sandia Pulse Reactor SPR-III pulses show very good agreement for a full range of pulse magnitudes. The model is also applied to Z-pinch externally driven neutron assembly (ZEDNA) type reactor designs to model both normal and off-normal ZEDNA operations.

  12. State of the art and prospects of fast neutron reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zrodnikov, A.V.; Mittenkov, F.M.; Poplavsky, V.M.; Kiryushin, A.I. [Physics and Power Eng. Inst., Obninsk (Russian Federation). State Sci. Centre

    1997-10-01

    On the basis of experience of fast reactor design, construction and operation gained in Russia, this paper outlines their state of the art. The high maturity and efficiency of this type of nuclear power development in Russia and the equalization of the economic characteristics of thermal and fast reactors is shown, as well as the expediency of improvement of nuclear power environmental characteristics owing to fast reactors incorporation. (orig.) 7 refs.

  13. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    Science.gov (United States)

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system.

  14. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

  15. Terrestrial and Reactor Antineutrinos in Borexino

    Science.gov (United States)

    Chen, M. C.; Calaprice, F. P.; Rothschild, C. G.

    1998-10-01

    The Earth is an abundant source of antineutrinos coming from the decay of radioactive elements in the mantle and crust. Detecting these antineutrinos is a challenge due to their small cross section and low energies. The Borexino solar neutrino experiment will also be an excellent detector for barν_e. With 300 tons of ultra-low-background liquid scintillator, surrounded by an efficient muon veto, the inverse-β-decay reaction: barνe + p arrow e^+ + n (Q = 1.8 MeV), can be exploited to detect terrestrial antineutrinos from the uranium and thorium decay chains, with little background. A direct measurement of the total uranium and thorium abundance would establish important geophysical constraints on the heat generation and thermal history of the Earth. Starting with the most recent uranium and thorium distribution and abundance data, and employing a global map of crustal type and thickness, we calculated the antineutrino fluxes for several sites. We estimate a terrestrial antineutrino event rate in Borexino of 10 events per year. This small signal can be distinguished over the neutrino background from the world's nuclear power reactors by measuring the positron energy spectrum from the barνe events. The possibility to perform a long-baseline oscillation experiment, reaching Δ m^2 ≈ 10-6 eV^2, using the nuclear reactors in Europe will also be discussed.

  16. Pebble Bed Reactor Plant screening evaluation. Volume 1. Overall plant and reactor system

    International Nuclear Information System (INIS)

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW/sub t/ Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system. Core scoping studies were performed which evaluated the effects of annular and cyclindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations

  17. Innovation and research in reactor safety

    International Nuclear Information System (INIS)

    In line with the engineered safeguards principle of in-depth safety, the survey article deals with innovation and research in the field of reactor safety, improvements in plant operation, innovation in accident management, and reduction of the consequences of severe accidents. The survey reveals that the development and application of innovative and efficient technologies is aimed primarily at the management of aging and of the operating life, and at simplifying and improving operations processes. Another area of innovation is accident management. In this respect, some of the main areas under development are the expansion of the multi-level safety concept, the introduction of further accident control measures so as to complete the spectrum of accidents covered, the quantification of safety margins by means of the application of modern methods of computation, and the introduction of passive elements reducing the need for fast countermeasures to be initiated by the plant operating personnel. The authors conclude that, on the whole, light water reactors attain a level of safety which, in combination with corresponding efforts in the economic sector, is a precondition for the renaissance of nuclear technology in the century just begun. The second part of the article, which is to be published in July, will deal mainly with the reduction of consequences of severe accidents. (orig.)

  18. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    International Nuclear Information System (INIS)

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can be further increased either by the concept of heterogeneous cores or by using carbide or nitride fuel (breeding gain about0.35). Recently, the old idea of breeding in advanced pressurized water reactors (PWRs) has been taken up again with the objective of attaining a gain of 0.05. Unfortunately, these objectives had to be limited to a conversion ratio of 0.9 for safety reasons, and it is not certain whether operation will be rewarding economically. The strategy of substituting PWRs is examined using the French example. By gradually introducing LMFBRs, the cumulated uranium supplies in France can be kept within reasonable limits, which means that they attain three to four times the home resources. This is not possible with advanced LWRs, which can be considered only as a possible backup solution for plutonium recycling into PWRs

  19. GIF's activities and safety approaches for Generation IV reactors

    International Nuclear Information System (INIS)

    Safety features, challenges and approaches for the Generation IV reactor systems are outlined on the promising six nuclear energy systems such as the Sodium-cooled Fast Reactor (SFR), the Very High Temperature Reactor (VHTR), the Lead-cooled Fast Reactor (LFR), the Supercritical Water-cooled Reactor (SCWR), the Gas-cooled Fast Reactor (GFR), and the Molten Salt Reactor (MSR) for further development. (J.P.N.)

  20. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

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

  1. Flow Simulation and Optimization of Plasma Reactors for Coal Gasification

    Institute of Scientific and Technical Information of China (English)

    冀春俊; 张英姿; 马腾才

    2003-01-01

    This paper reports a 3-d numerical simulation system to analyze the complicatedflow in plasma reactors for coal gasification, which involve complex chemical reaction, two-phaseflow and plasma effect. On the basis of analytic results, the distribution of the density, tempera-ture and components' concentration are obtained and a different plasma reactor configuration isproposed to optimize the flow parameters. The numerical simulation results show an improvedconversion ratio of the coal gasification. Different kinds of chemical reaction models are used tosimulate the complex flow inside the reactor. It can be concluded that the numerical simulationsystem can be very useful for the design and optimization of the plasma reactor.

  2. Nuclear reactor safety and Federal regulation

    International Nuclear Information System (INIS)

    Public confidence in nuclear reactors requires that technical people translate complex safety information into a form that the public can understand well enough to make a judgment. An overall picture is drawn of the major areas of concern: (1) risks and safety measures, (2) government regulation, (3) licensing, (4) plant operation, (5) safety experience, and (6) quality assurance. Although the possibilities of a reactor core melting through the concrete containment barrier are slight, rigorous safety efforts are required. Government regulation and technical developments have developed concurrently so that the high standards set for government facilities can be carried over to commercial efforts. There are two stages in the licensing procedure: a construction permit and an operating license. Reviews of the proposed site, design, emergency cooling systems are all held, followed by a public hearing. Inspection and backfitting of new safety equipment are required in operating plants. The 60 plants now in operation have a good performance record, but good management for quality assurance increases safety and efficiency factors

  3. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  4. Accessibility and Radioactivity Calculations for Nuclear Reactor Shutdown System

    International Nuclear Information System (INIS)

    An important consideration in the design of power reactors is providing access to the reactor cooling system for the purposes of maintenance, repair and refuelling. The major sources of radiation which tend to prohibit such access are: induced activity of the reactor coolant, activated impurities in the reactor coolant and radiation originating in the reactor core both during reactor operation and after shut down. Impurities in the reactor coolant may be present in high enough concentrations so that their activation restricts accessibility for maintenance after shutdown. When water being used as a coolant, the activity of the water itself is very short- lived but their corrosive nature, resultant high impurity and induced activity of structural material are the major source of activity in the system after reactor shutdown. In this case, it may be necessary to chemically remove some of the impurity by a purification process to prevent a build up of long-lived induced activity in the system from restricting access to the plant, and to keep the radiation dose at the working places within the permissible limits. A mathematical modelling is developed. A system of coupled first-order linear differential equations describing adequately the activity behaviour has to be derived and solved. It treats the determination of equilibrium concentrations of impurities on system surface , and the effect of release of fission products from the reactor core

  5. Method of surface treatment for structure and facility in reactor

    International Nuclear Information System (INIS)

    Surfaces of weld zones, in contact with liquid, of structures and equipments in a reactor made of austenite stainless steels disposed in the reactor water of a reactor pressure vessel are melted by laser. Then, heat affected zones and grain boundary segregation portions, etc. with low corrosion resistance formed under irradiation are melted by laser beams and the molten surfaces are quenched by the surrounding reactor water. In this case, ferrites are formed to provide a two-phase structure. This can improve the corrosion resistance. Further, plasma technology can be used instead of the laser method. (I.S.)

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

    International Nuclear Information System (INIS)

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

  7. Contributions of research Reactors in science and technology

    International Nuclear Information System (INIS)

    In the present paper, after defining a research reactor, its basic constituents, types of reactors, their distribution in the world, some typical examples of their uses are given. Particular emphasis in placed on the contribution of PARR-I (Pakistan Research Reactor-I), the 5 MW Swimming Pool Research reactor which first became critical at the Pakistan Institute of Nuclear Science and Technology (PINSTECH) in Dec. 1965 and attained its full power in June 1966. This is still the major research facility at PINSTECH for research and development. (author)

  8. Modelling and control design for SHARON/Anammox reactor sequence

    OpenAIRE

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation...

  9. Chernobyl and the safety of nuclear reactors in OECD countries

    International Nuclear Information System (INIS)

    This report assesses the possible bearing of the Chernobyl accident on the safety of nuclear reactors in OECD countries. It discusses analyses of the accident performed in several countries as well as improvements to the safety of RBMK reactors announced by the USSR. Several remaining questions are identified. The report compares RBMK safety features with those of commercial reactors in OECD countries and evaluates a number of issues raised by the Chernobyl accident

  10. Remote controlled stud bolt handling device for reactor pressure vessel

    International Nuclear Information System (INIS)

    In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

  11. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    Directory of Open Access Journals (Sweden)

    José Tavares de Sousa

    2008-01-01

    Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

  12. Gas-liquid separation mechanism and reactor

    International Nuclear Information System (INIS)

    A space above the reactor core mounted in a pressure vessel of a BWR type reactor is surrounded with a cylindrical shroud. Since vortex and stagnation of coolants are liable to be formed over the entire periphery at the upper portion of the shroud, inlets through which steam bubbles enter are opened to the entire circumference along the upper portion of the shroud. This can attain the gas-liquid separation effect from the entire periphery thereof. Further, since the coolant flowing vector changes abruptly at a portion where the coolant flow turns from the upward to the downward direction, steam bubbles are swirled or stagnated on the side of the downward flow channel in that portion. In view of the above, a coolant inlet submerged in the coolant is disposed to that portion, so that the steam bubbles are captured, involved in the subsequently flowing coolants and urged upwardly into the space above the coolant surface, thereby maintaining extremely high spontaneous recycling force and recycling amount. (T.M.)

  13. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

    In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

  14. Combined Reactor and Microelectrode Measurements in Laboratory Grown Biofilms

    DEFF Research Database (Denmark)

    Larsen, Tove; Harremoës, Poul

    1994-01-01

    were carried out with aerobic glucose and starch degrading biofilms. The well described aerobic glucose degradation biofilm system was used to test the combined reactor set-up. Results predicted from known biofilm kinetics were obtained. In the starch degrading biofilm, basic assumptions were tested......A combined biofilm reactor-/microelectrode experimental set-up has been constructed, allowing for simultaneous reactor mass balances and measurements of concentration profiles within the biofilm. The system consists of an annular biofilm reactor equipped with an oxygen microelectrode. Experiments...

  15. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  16. Fiber-Optical Sensors: Basics and Applications in Multiphase Reactors

    OpenAIRE

    Guozheng Li; Shifang Yang; Chao Yang; Xiangyang Li

    2012-01-01

    This work presents a brief introduction on the basics of fiber-optical sensors and an overview focused on the applications to measurements in multiphase reactors. The most commonly principle utilized is laser back scattering, which is also the foundation for almost all current probes used in multiphase reactors. The fiber-optical probe techniques in two-phase reactors are more developed than those in three-phase reactors. There are many studies on the measurement of gas holdup using fiber-opt...

  17. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    International Nuclear Information System (INIS)

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I ampersand C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models

  18. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  19. Spent fuel situation at the ASTRA Seibersdorf and the TRIGA Vienna research reactors

    International Nuclear Information System (INIS)

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atomic Institut Vienna and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down on July 31st, 1999 and decommissioning of the ASTRA reactor and the shut down of the ARGONAUT reactor Graz on July 31, 2004 only the TRIGA reactor remains operational. The MTR fuel elements of the ASTRA reactor have been shipped in spring 2001 to Savannah River and the fuel plates from the ARGONAUT reactor Graz in December 2005 under the DOE fuel return programme. (author)

  20. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  1. Investigation of small and modular-sized fast reactor

    International Nuclear Information System (INIS)

    In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

  2. Component and operation experience of reactor TRIGA MARK II

    International Nuclear Information System (INIS)

    Reactor TRIGA MARK II is Jozef Stefan Institute's research reactor. It has been operating since 1966. A probabilistic approach of reactor safety estimation was used first in 1989 when a Probabilistic Safety Analysis (PSA) of the reactor was performed. A lack of reactor component data was found as the major problem in probabilistic assessment. It was decided to continue the work with specific data base development. The project has been divided in two phases. In the first phase specific data from year 1985 to 1990 were collected. In the second phase the collected data were treated. The comparison of generic and specific data showed significant difference between the generic and specific data and leads to a conclusion that a generic data based PSA has a limited credibility indicating that there is a need to build a specific data base for research reactors. The TRIGA MARK II research reactor has three major purposes: operator training, research involving neutrons and isotope production. The paper represents specific data base formation for TRIGA MARK II research reactor in Podgorica. Specific data on reactor scrams, components operation and human errors were collected. The data of fifteen components were estimated by classical and Bayesian method. The results of both methods are very different. Because of good specific data the results of classical methods were preferred. The comparison of specific and generic data showed that there is a great need to build a specific data base for research reactors. It is expected to use the specific data for existing PSA of TRIGA MARK II reactor reevaluation and optimisation of its operation. (authors)

  3. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  4. Nuclei in nuclear reactors and stars

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, Kazuhiro [Nagoya Univ. (Japan). School of Engineering

    1997-03-01

    To evaluate the characteristic properties of fission products, the fission yield and decay data (decay constant, decay energy and release rate of decayed neutron) of about 1000 kinds of nuclide are necessary. From 1970 to 1980 years, the decay heat and the characteristics of delayed neutrons for the principle fission nuclide such as {sup 235}U, {sup 238}U and {sup 239}Pu were appreciable to use in the general light water reactor. We discussed relation between calculation of the characteristics of fission products and study of unstable nuclide and astronomical physics. Total calculation of decay heat was investigated by formal solution. To increase the evaluation values, the accuracy of fission yield and the decay data of nuclides with short life should be increased by measurements. (S.Y.)

  5. Reactor - and accelerator-based filtered beams

    International Nuclear Information System (INIS)

    The neutrons produced in high flux nuclear reactors and in accelerator, induced fission and spallation reactions, represent the most intense sources of neutrons available for research. However, the neutrons from these sources are not monoenergetic, covering the broad range extending from 10-3 eV up to 107 eV or so. In order to make quantitative measurements of the effects of neutrons and their dependence on neutron energy it is desirable to have mono-energetic neutron sources. The paper describes briefly methods of obtaining mono-energetic neutrons and different methods of filtration. This is followed by more detailed discussion of neutron window filters and a summary of the filtered beam facilities using this technique. The review concludes with a discussion of the main applications of filtered beams and their present and future importance

  6. Experience, status and future of the computerized reactor instrumentation at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    The paper describes the 33 years old history of the instrumentation of the TRIGA reactor Vienna and focuses on the present computerized instrumentation installed in 1992. The experience of three years of operation is discussed and some of the failures are analyzed. Potential future problems both with soft- and hardware as well as with spare part supplies are analyzed. (author). 6 figs

  7. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  8. Uncertainties and reliability theories for reactor safety

    International Nuclear Information System (INIS)

    This paper presents: (i) a classification of uncertainties and of reliability models for reactor safety; (ii) a general methodology to include these uncertainties into reliability analysis; (iii) a discussion about the relative advantages and the limitations of various reliability theories (specifically, of inductive and deductive, parametric and nonparametric, second-moment and full-distribution theories). For example, it is shown that second-moment theories, which were originally suggested to cope with the scarcity of data, and which have been proposed recently for the safety analysis of secondary containment vessels, are the least capable of incorporating statistical uncertainty. In the paper, the focus is on reliability models for external threats, due to the reasons mentioned above. As an application example, the effect of statistical uncertainty on seismic risk is studied using parametric full-distribution models. (Auth.)

  9. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  10. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  11. Current tendencies and perspectives of development research reactors of Russia

    International Nuclear Information System (INIS)

    Full text: During more than fifty years many Research Reactors were constructed under Russian projects, and that is a considerable contribution to the world reactor building. The designs of Research Reactors, constructed under Russian projects, appeared to be so successful, that permitted to raise capacity and widen the range of their application. The majority of Russian Research Reactors being middle-aged are far from having their designed resources exhausted and are kept on the intensive run still. In 2000 'Strategy of nuclear power development in Russia in the first half of XXI century' was elaborated and approved. The national nuclear power requirements and possible ways of its development determined in this document demanded to analyze the state of the research reactors base. The analysis results are presented in this report. The main conclusion consists in the following statement: on the one hand quantity and experimental potentialities of domestic Research Reactors are sufficient for the solution of reactor materials science tasks, and on the other hand the reconstruction and modernization appears to be the most preferable way of research reactors development for the near-term outlook. At present time the modernization and reconstruction works and works on extension of operational life of high-powered multipurpose MIR-M1, SM-3, IRV-1M, BOR-60, IVV-2M and others are conducted. There is support for the development of Research Reactors, intended for carrying out the fundamental investigations on the neutron beams. Toward this end the Government of Russia gives financial and professional support with a view to complete the reactor PIK construction in PINPh and the reactor IBR-2 modernization in JINR. In future prospect Research Reactors branch in Russia is to acquire the following trends: - limited number of existent scientific centers, based on the construction sites, with high flux materials testing research reactors, equipped with experimental facilities

  12. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  13. Small reactors in the Canadian context: opportunities and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Walker, R.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This presentation discusses the opportunities and challenges for small reactors in Canada. It concludes by suggesting that the success of small reactors in Canada will depend on a number of factors including private sector investment, access to international markets, stable, equitable and adaptable regulatory regime, public trust and technology.

  14. Cars applications in chemical reactors, combustion and heat transfer

    Science.gov (United States)

    Greenhalgh, D. A.; Porter, F. M.

    1986-08-01

    This paper illustrates the use of the CARS technique in the fields of Chemical Reactor engineering, combustion and Heat Transfer. Examples of recent results from a catalytic chemical reactor, an operating production petrol engine and an oil spray furnace are given. The experimentally determined accuracy of CARS nitrogen thermometry for both mean and single pulse measurements is presented.

  15. DNA-Based Enzyme Reactors and Systems

    Directory of Open Access Journals (Sweden)

    Veikko Linko

    2016-07-01

    Full Text Available During recent years, the possibility to create custom biocompatible nanoshapes using DNA as a building material has rapidly emerged. Further, these rationally designed DNA structures could be exploited in positioning pivotal molecules, such as enzymes, with nanometer-level precision. This feature could be used in the fabrication of artificial biochemical machinery that is able to mimic the complex reactions found in living cells. Currently, DNA-enzyme hybrids can be used to control (multi-enzyme cascade reactions and to regulate the enzyme functions and the reaction pathways. Moreover, sophisticated DNA structures can be utilized in encapsulating active enzymes and delivering the molecular cargo into cells. In this review, we focus on the latest enzyme systems based on novel DNA nanostructures: enzyme reactors, regulatory devices and carriers that can find uses in various biotechnological and nanomedical applications.

  16. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  17. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  18. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  19. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  20. Recommended safety objectives, principles and requirements for mini-reactors

    International Nuclear Information System (INIS)

    Canadian and international publications containing objectives, principles and requirements for the safety of nuclear facilities in general and nuclear power plants in particular have been reviewed for their relevance to mini-reactors. Most of the individual recommendations, sometimes with minor wording changes, are applicable to mini-reactors. However, some prescriptive requirements for the shutdown, emergency core cooling and containment systems of power reactors are considered inappropriate for mini-reactors. The Advisory Committee on Nuclear Safety favours a generally non-prescriptive approach whereby the applicant for a mini-reactor license is free to propose any means of satisfying the fundamental objectives, but must convince the regulatory agency to that effect. To do so, a probabilistic safety assessment (PSA) would be the favoured procedure. A generic PSA for all mini-reactors of the same design would be acceptable. Notwithstanding this non-prescriptive approach, the ACNS considers that it would be prudent to require the existence of at least one independent shutdown system and two physically independent locations from which the reactor can be shut down and the shutdown condition monitored, and to require provision for an assumed loss of integrity of the primary cooling system's boundary unless convincing arguments to the contrary are presented. The ACNS endorses in general the objectives and fundamental principles proposed by the interorganizational Small Reactor Criteria working group, and intends to review and comment on the documents on specific applications to be issued by that working group

  1. Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

    International Nuclear Information System (INIS)

    Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

  2. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  3. A Joint Report on PSA for New and Advanced Reactors

    International Nuclear Information System (INIS)

    This report addresses the application of Probabilistic Safety Assessment (PSA) to new and advanced nuclear reactors. As far as advanced reactors are concerned, the objectives were to characterize the ability of current PSA technology to address key questions regarding the development, acceptance and licensing of advanced reactor designs, to characterize the potential value of advanced PSA methods and tools for application to advanced reactors, and to develop recommendations for any needed developments regarding PSA for these reactors. As far as the design and commissioning of new nuclear power plants is concerned, the objectives were to identify and characterize current practices regarding the role of PSA, to identify key technical issues regarding PSA, lessons learned and issues requiring further work; to develop recommendations regarding the use of PSA, and to identify future international cooperative work on the identified issues. In order to reach these objectives, questionnaires had been sent to participating countries and organisations

  4. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  5. Development, utilization, and future prospects of materials test reactors

    International Nuclear Information System (INIS)

    Reactor radiation affects the chemical and physical properties of materials. These changes can be very drastic in certain cases. Special test reactors have therefore been built since the 1950's and specific skills were developed to expose materials specimens to the precise irradiation conditions required. Materials testing reactors are those research reactor facilities which are designed and operated predominantly for studies into radiation damage. About a dozen plants in European communities (EC) Member States and in the US can be identified in this category, with 5 to 100 MW fission power and neutron fluxes between 5 x 1013 and 1015 cm-2s-1. The paper elaborates common aspects of development, utilization, and future prospects of US and EC materials testing reactors, and indicates the most significant differences

  6. Analysis of Reactor Deployment Scenarios with Introduction of SFR Breakeven Reactors and Burners Using DANESS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2008-01-15

    Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.

  7. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  8. Nuclear reactors transients identification and classification system

    International Nuclear Information System (INIS)

    This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

  9. Status and future plan of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Agency (JAEA) is a light water cooling tank typed reactor. JMTR has been used for fuel and material irradiation studies for LWRs, HTGR, fusion reactor and RI production. Since the JMTR is connected with hot laboratory through the canal, re-irradiation tests can conduct easily by safety and quick transportation of irradiation samples. First criticality was achieved in March 1968, and operation was stopped from August, 2006 for the refurbishment. The reactor facilities are refurbished during four years from the beginning of FY 2007, and necessary examination and work are carrying out on schedule. The renewed and upgraded JMTR will start from FY 2011 and operate for a period of about 20 years (until around FY 2030). The usability improvement of the JMTR, such as higher reactor available factor, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussing as the preparations for re-operation. (author)

  10. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  11. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  12. Developing research reactor coalitions and centres of excellence

    International Nuclear Information System (INIS)

    Research reactors continue to play a key role in the development of peaceful uses of atomic energy. They are used for a variety of purposes such as education and training, production of medical and industrial isotopes, non-destructive testing, analytical studies, modification of materials, for research in physics, biology and materials science, and in support of nuclear power programmes. The IAEA Research Reactor Data Base lists about 250 operational research reactors worldwide, many of which have been operating for more than 40 years. Through both statistical and anecdotal evidence, it is clear that many of these reactors are under utilized, face critical issues related to sustainability, and must make important decisions concerning future operation. These challenges are occurring in the context of increased concerns over global non-proliferation and nuclear material security, due to which research reactor operators are coming under increased pressure to substantially improve physical security and convert to the use of low enriched uranium (LEU) fuel. Thus, there is a complex environment for research reactors, and one in which underutilized and therefore likely poorly funded facilities invoke particular concern. any research reactors are challenged to generate sufficient income to offset operational costs, often in a context of declining political and/or public support. Many research reactor operators have limited access to potential customers for their services and are not familiar with the business planning concepts needed to secure additional commercial revenues or governmental or international programme funding. This not only results in reduced income for the facilities involved, but sometimes also in research reactor services priced below full cost, preventing recovery of back-end costs and creating unsustainable market norms. Parochial attitudes and competitive behaviour restrict information sharing, dissemination of best practices, and mutual support that

  13. American Nuclear Society standards for TRIGA reactors and their use

    International Nuclear Information System (INIS)

    The American Nuclear Society established a committee (ANS-15) with the expressed charter to develop standards for research reactors. These standards were to cover all aspects of research reactor operations, maintenance and administration. Standards have been written in every area of research reactor operations that the research reactor community has deemed important. One of the uppermost goals of the Standards Committee work is to produce standards that provide guidance and help to the research reactor community in a timely manner. To make the standards meaningful requires a great deal of cooperation between all segments of the reactor community. The research reactors - whether they are private, university or government owned - have a mission to perform. At the same time, the regulatory agencies also have a mission to perform, and with a spirit of mutual respect and cooperation, both can accomplish their goals. In the last five years this spirit has been present, and a number of very good standards have resulted. These standards should be a part of every research reactor library. In particular ANS-15.16 and ANS-15.1 have been endorsed by the regulatory agencies and are being used to evaluate submittals

  14. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  15. Nuclear power reactors and hydrogen storage systems

    International Nuclear Information System (INIS)

    Among conclusions and results come by, a nuclear-electric-hydrogen integrated power system was suggested as a way to prevent the energy crisis. It was shown that the hydrogen power system using nuclear power as a leading energy resource would hold an advantage in the current international situation as well as for the long-term future. Results reported provide designers of integrated nuclear-electric-hydrogen systems with computation models and routines which will allow them to explore the optimal solution in coupling power reactors to hydrogen producing systems, taking into account the specific characters of hydrogen storage systems. The models were meant for average computers of a type easily available in developing countries. (author)

  16. Deposition reactors for solar grade silicon: A comparative thermal analysis of a Siemens reactor and a fluidized bed reactor

    Science.gov (United States)

    Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.

    2015-12-01

    Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.

  17. Non destructive examination of Reactor DR-3. Reactor wall, horisontal experimental tubes, up- and down comers

    International Nuclear Information System (INIS)

    The initial scope of work was to perform thickness/corrosion measurements of one up-comer and one down-comer, perform thickness/corrosion measurements in selected areas of the reactor wall and horizontal experimental pipes inside the reactor. Furthermore the lower circumferential weld and the connected longitudinal weld should be inspected to the extent possible, without major changes of the manipulator. Eddy current was performed in the same areas. Also hardness tests were carried out in four positions inside the reactor. Due to the outcome of the above examinations, additional metallurgical and dye penetrant examinations (PT) were carried out. The examination of the up- and down comers showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The examination of the horizontal tubes showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The hardness test showed increased hardness compared to calibration values. The examination of the reactor wall base material revealed several indications located in different depths in the plate. Some indications have been proved to be connected to the inner surface, while most indications appear to be either inclusions or areas corroded from the outside reactor wall. Minimum measured wall thickness is between 4.2 and 11.0 mm. There is, however, no evidence that these values are caused by corrosion at the outer reactor surface. The ET showed no signs of service induced cracks. The hardness test showed values close to calibration values. The extensive number of indications has resulted in additional

  18. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  19. Research and development on next generation reactor (phase I)

    International Nuclear Information System (INIS)

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author)

  20. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  1. Autothermal reactor concept for combined oxidative coupling and methane reforming

    NARCIS (Netherlands)

    Tiemersma, T.P.; Sint Annaland, van M.; Kuipers, J.A.M.

    2009-01-01

    A novel autothermal reactor concept has been developed for the simultaneous production of ethylene by oxidative coupling (OCM) and synthesis gas by steam reforming of methane (SRM), supported by a detailed numerical modeling study. The proposed reactor consists of two separate reaction chambers whic

  2. High-temperature membrane reactors: potential and problems

    NARCIS (Netherlands)

    Saracco, G.; Neomagus, H.W.J.P.; Versteeg, G.F.; Swaaij, van W.P.M.

    1999-01-01

    The most recent literature in the field of membrane reactors is reviewed, four years after an analogous effort of ours (Saracco et al., 1994), describing shortly the potentials of these reactors, which now seem to be well established, and focusing mostly on problems towards practical exploitation. S

  3. High-temperature membrane reactors : potential and problems

    NARCIS (Netherlands)

    Saracco, G.; Neomagus, H.W.J.P.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    The most recent literature in the field of membrane reactors is reviewed, four years after an analogous effort of ours, describing shortly the potentials of these reactors, which now seem to be well established, and focusing mostly on problems towards practical exploitation. Since then, progress has

  4. Refurbishment and Modernisation of PUSPATI TRIGA Reactor and Lessons Learnt

    International Nuclear Information System (INIS)

    The PUSPATI TRIGA Reactor first became critical in June 1982, and has been in operation since then. Over the years, several of the reactor systems, structures and components (SSCs) experience ageing and obsolescence problems and had to be refurbished, replaced or modernised. Initially refurbishment or replacements were carried out with SSCs of equivalent quality or capability. Subsequently SSCs were replaced with higher specification to allow for future upgrading of the reactor. Features of new SSCs should include all features of SSCs to be replaced and consider human machine interface to avoid any incidents. Lessons learnt over the years have been applied to the reactor control console modernisation project. In this project the involvement of our personnel during the design, fabrication and testing stages will enable us to have the capability to solve any associated problems with minimal vendor involvement. The close cooperation between regulators of Malaysia and vendor country was also beneficial to ensure that the project meet international safety standards

  5. Pacific Northwest Laboratory Monthly Activities Report APRIL 1966 on AEC Division of Reactor Development and Technology

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Fawcett

    1966-05-01

    This report has the following sections: Summary of Activities; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

  6. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  7. Incorporation of statistical distribution of particle properties in chemical reactor design and operation: the cooled tubular reactor

    OpenAIRE

    Wijngaarden, R.J.; Westerterp, K.R.

    1992-01-01

    Pellet heat and mass transfer coefficients inside packed beds do not have definite deterministic values, but are stochastic quantities with a certain distribution. Here, a method is presented to incorporate the stochastic distribution of pellet properties in reactor design and operation models. The theory presented is illustrated with a number of examples. It is shown that pellet-scale statistics have an impact on cooled tubular reactor design and operation. Cooled tubular reactor design is d...

  8. Nitrification of industrial and domestic saline wastewaters in moving bed biofilm reactor and sequencing batch reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bassin, Joao P. [Programa de Engenharia Quimica/COPPE, Universidade Federal do Rio de Janeiro, Centro de Tecnologia, Bloco G - sala 116, P.O. Box 68502, 21941-972 Rio de Janeiro, RJ (Brazil); Dezotti, Marcia, E-mail: mdezotti@peq.coppe.ufrj.br [Programa de Engenharia Quimica/COPPE, Universidade Federal do Rio de Janeiro, Centro de Tecnologia, Bloco G - sala 116, P.O. Box 68502, 21941-972 Rio de Janeiro, RJ (Brazil); Sant' Anna, Geraldo L. [Programa de Engenharia Quimica/COPPE, Universidade Federal do Rio de Janeiro, Centro de Tecnologia, Bloco G - sala 116, P.O. Box 68502, 21941-972 Rio de Janeiro, RJ (Brazil)

    2011-01-15

    Nitrification of saline wastewaters was investigated in bench-scale moving-bed biofilm reactors (MBBR). Wastewater from a chemical industry and domestic sewage, both treated by the activated sludge process, were fed to moving-bed reactors. The industrial wastewater contained 8000 mg Cl{sup -}/L and the salinity of the treated sewage was gradually increased until that level. Residual substances present in the treated industrial wastewater had a strong inhibitory effect on the nitrification process. Assays to determine inhibitory effects were performed with the industrial wastewater, which was submitted to ozonation and carbon adsorption pretreatments. The latter treatment was effective for dissolved organic carbon (DOC) removal and improved nitrification efficiency. Nitrification percentage of the treated domestic sewage was higher than 90% for all tested chloride concentrations up to 8000 mg/L. Results obtained in a sequencing batch reactor (SBR) were consistent with those attained in the MBBR systems, allowing tertiary nitrification and providing adequate conditions for adaptation of nitrifying microorganisms even under stressing and inhibitory conditions.

  9. Nitrification of industrial and domestic saline wastewaters in moving bed biofilm reactor and sequencing batch reactor.

    Science.gov (United States)

    Bassin, João P; Dezotti, Marcia; Sant'anna, Geraldo L

    2011-01-15

    Nitrification of saline wastewaters was investigated in bench-scale moving-bed biofilm reactors (MBBR). Wastewater from a chemical industry and domestic sewage, both treated by the activated sludge process, were fed to moving-bed reactors. The industrial wastewater contained 8000 mg Cl(-)/L and the salinity of the treated sewage was gradually increased until that level. Residual substances present in the treated industrial wastewater had a strong inhibitory effect on the nitrification process. Assays to determine inhibitory effects were performed with the industrial wastewater, which was submitted to ozonation and carbon adsorption pretreatments. The latter treatment was effective for dissolved organic carbon (DOC) removal and improved nitrification efficiency. Nitrification percentage of the treated domestic sewage was higher than 90% for all tested chloride concentrations up to 8000 mg/L. Results obtained in a sequencing batch reactor (SBR) were consistent with those attained in the MBBR systems, allowing tertiary nitrification and providing adequate conditions for adaptation of nitrifying microorganisms even under stressing and inhibitory conditions.

  10. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    Science.gov (United States)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  11. Power instability and stochastic dynamics of periodic pulsed reactors

    International Nuclear Information System (INIS)

    This paper reports that physicists dealing with conventional reactor dynamics recognize two types of instability and reactor behavior beyond the stability region: asymptotic excursions and nonlinear periodic oscillations. A periodically pulsed reactor (PPR) has another peculiar instability: Under certain conditions, its power tends to oscillate at a frequency just twice less than the reactor pulsation frequency. The PPR dynamics far beyond the stability region are analyzed by using a discrete nonlinear model. A PPR with a negative temperature reactivity effect inevitably shows the chaotic power pulse energy behavior known as deterministic chaos. The way by which a reactor goes to chaos is defined by the time dependence of the feedback and by the kind of dynamics model used

  12. Status of innovative small and medium sized reactor designs 2005. Reactors with conventional refuelling schemes

    International Nuclear Information System (INIS)

    There is a renewed interest in Member States in the development and application of small and medium sized reactors (SMRs). In the near term, most new NPPs are likely to be evolutionary designs building on proven systems while incorporating technological advances and often the economics of scale, resulting from the reactor outputs of up to 1600 MW(e). For the longer term, the focus is on innovative designs aiming to provide increased benefits in the areas of safety and security, non-proliferation, waste management, resource utilization and economy, as well as to offer a variety of energy products and flexibility in design, siting and fuel cycle options. Many innovative designs are reactors within the small-to-medium size range, having an equivalent electric power less than 700 MW(e) or even less than 300 MW(e). The projected timelines of readiness for deployment are generally between 2010 and 2030. The objective of this report is to provide Member States, including those just considering the initiation of nuclear power programmes, and those already having practical experience in nuclear power, with a balanced and objective information on important development trends and objectives of innovative SMRs for a variety of uses, on the achieved state-of-the-art in design and technology development for such reactors and on their design and regulatory status. The report is intended for many categories of stakeholders, including regulators, electricity producers, designers, non-electrical producers and policy makers. The main chapters of this report, addressed to all abovementioned groups of stakeholders, provide a summary of major specifications, applications and user-related special features of innovative SMRs, outline the achieved design and regulatory status and its progress since previous IAEA publications, review targeted deployment dates, fuel cycle options, design approaches used to meet design objectives in specific subject areas, enabling technologies and current

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

    International Nuclear Information System (INIS)

    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.)

  14. Research reactors

    International Nuclear Information System (INIS)

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  15. Recycle strategies for fast reactors and related fuel cycle technologies

    International Nuclear Information System (INIS)

    Full text: 1. Introduction Fast reactors and related fuel cycle (hereinafter referred to as 'Fast reactor cycle') technologies have the potential of contributing to long-term energy security due to effective use of uranium and plutonium resources, and reduction of the heat generation and potential toxicity of high-level radioactive wastes by burning long-lived minor actinides (MA) recovered from spent fuels of light-water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non- proliferation as addressed in GNEP (Global Nuclear Energy Partnership). With these features, R and Ds toward their commercialization have been promoted vigorously and globally as a future vision of nuclear energy. 2. Recycle strategies in each country In Japan, it is determined that after burning uranium in light water reactors, plutonium is recovered from spent fuel and used for light water reactors at the moment and for fast reactors in the future. In order to make it possible, Fast Reactor Cycle Technology Development (FaCT) Project has been promoted with a combination of oxide-fueled sodium-cooled reactors, advanced aqueous reprocessing, and simplified pelletizing fuel fabrication adopted as a main concept aiming at startup of a demonstration reactor around 2025 and commercialization before around 2050. In France, a comparison of the basic specifications between an oxide-fueled sodium-cooled reactor and a carbide (or nitride)-fueled gas-cooled reactor has currently been promoted towards technological selection for a prototype reactor in 2012 in accordance with 'The 2006 planning act on the sustainable management of radioactive materials and waste (Act 2006- 739)' enacted in 2006. Based on the results, France aims at startup of the prototype reactor in 2020 and commercialization in around 2040. For reprocessing, methods which extract actinides collectively such as GANEX has been developed to enhance proliferation resistance

  16. 75 FR 54657 - University of Florida; University of Florida Training Reactor; Environmental Assessment and...

    Science.gov (United States)

    2010-09-08

    ... COMMISSION University of Florida; University of Florida Training Reactor; Environmental Assessment and... a significant environmental impact. I. Radiological Impact Environmental Effects of Reactor... licensee), which would authorize continued operation of the University of Florida Training Reactor...

  17. Solution structure of loperamide and -cyclodextrin inclusion complexes using NMR spectroscopy

    Indian Academy of Sciences (India)

    Santosh Kumar Upadhyay; Syed Mashhood Ali

    2009-07-01

    Loperamide (LPR) is a synthetic, poorly water soluble, peripherally acting opiate agonist drug used for the treatment of diarrhea. Major challenges in formulating this drug for clinical applications include solubility enhancement and improved stability in biological systems. Cyclodextrins (CDs) are chiral, truncated cone shaped; cyclic oligosaccharides that can encapsulate a variety of poorly water soluble drug molecules into inclusion complexes, thereby increasing their stability and solubility. 1H NMR spectroscopic studies showed the inclusion complexation between -CD and LPR, based on the upfield shift changes in the -CD cavity protons (H-3' and H-5') and downfield shift changes in the guest (LPR) protons. 2D COSY spectral data was used for assignment of -CD as well as LPR protons and 2D ROESY spectral data to know the inclusion of LPR inside the -CD cavity. The 1 : 1 stoichiometry and overall association constant (a) were determined by using Scott’s plot method to be 68.805 M-1. 2D ROESY spectral data suggest that the inclusion of aromatic rings of LPR in -CD cavity can be from narrower as well as the wider rim side and the six possible 1 : 1 LPR : -CD inclusion complexes have been proposed. Thus, we anticipate that complexation of LPR with -CD would increase its solubility and stability in biological system.

  18. Linear regression and sensitivity analysis in nuclear reactor design

    International Nuclear Information System (INIS)

    Highlights: • Presented a benchmark for the applicability of linear regression to complex systems. • Applied linear regression to a nuclear reactor power system. • Performed neutronics, thermal–hydraulics, and energy conversion using Brayton’s cycle for the design of a GCFBR. • Performed detailed sensitivity analysis to a set of parameters in a nuclear reactor power system. • Modeled and developed reactor design using MCNP, regression using R, and thermal–hydraulics in Java. - Abstract: The paper presents a general strategy applicable for sensitivity analysis (SA), and uncertainity quantification analysis (UA) of parameters related to a nuclear reactor design. This work also validates the use of linear regression (LR) for predictive analysis in a nuclear reactor design. The analysis helps to determine the parameters on which a LR model can be fit for predictive analysis. For those parameters, a regression surface is created based on trial data and predictions are made using this surface. A general strategy of SA to determine and identify the influential parameters those affect the operation of the reactor is mentioned. Identification of design parameters and validation of linearity assumption for the application of LR of reactor design based on a set of tests is performed. The testing methods used to determine the behavior of the parameters can be used as a general strategy for UA, and SA of nuclear reactor models, and thermal hydraulics calculations. A design of a gas cooled fast breeder reactor (GCFBR), with thermal–hydraulics, and energy transfer has been used for the demonstration of this method. MCNP6 is used to simulate the GCFBR design, and perform the necessary criticality calculations. Java is used to build and run input samples, and to extract data from the output files of MCNP6, and R is used to perform regression analysis and other multivariate variance, and analysis of the collinearity of data

  19. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    In 2002, within the framework of the Department of Nuclear Energy's Technical Working Group on Fast Reactors (TWG-FR), and according to the expressed needs of the TWG-FR Member States to maintain and increase the present knowledge and expertise in fast reactor science and technology, the IAEA established its initiative seeking to establish a comprehensive, international inventory of fast reactor data and knowledge. More generally, at the IAEA meeting of senior officials convened to address issues of nuclear knowledge management underlying the safe and economic use of nuclear science and technology (Vienna, 17-19 June 2002), there was widespread agreement that, for sustainability reasons for fissile sources and waste management, long-term development of nuclear power as a part of the world's future energy mix will require the fast reactor technology. Furthermore, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This consensus concluded from the recognition of immediate need gave support to the IAEA initiative for fast reactor data and knowledge preservation. To implement the IAEA initiative, the scope of fast reactor knowledge preservation activities and a road map for implementation have been developed. The IAEA supports and coordinates data retrieval and interpretation efforts in the Member States joining the initiative and ensures the collaboration with other international organizations (mainly OECD/NEA) and eventually establishes and maintains a portal for accessing the fast reactor knowledge base. The IAEA assists Member State activities by providing an umbrella for information exchange and collaborative R and D to pool resources and expertise within the framework of the TWG-FR and the Agency's International Nuclear Information System (INIS) and Nuclear Knowledge Management Section (NKMS). The IAEA collects and summarizes the scientific and technical information

  20. Recycle Strategies for Fast Reactors and Related Fuel Cycle Technologies

    International Nuclear Information System (INIS)

    Fast reactors and related fuel cycle (hereafter referred to as 'fast reactor cycle') technologies have the potential to contribute to long term energy security owing to their effective use of uranium and plutonium resources, and to a reduction in the heat generation and potential toxicity of high level radioactive wastes by burning long lived minor actinides recovered from spent fuel from light water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non-proliferation as addressed in the Global Nuclear Energy Partnership. With these features, the research and development towards their commercialization has been promoted vigorously and globally as a future vision of nuclear energy. The introduction of fast reactor cycle systems will be carried out independently in each country according to its national conditions and nuclear energy policy. It should then be considered important to have a globally common consensus relating to safety philosophy, concepts of proliferation resistance, transuranic element burnup and recycling and so on. For the development and utilization of fast reactor cycle systems, while respecting each country's concept, it is essential to organize the technologies and concepts which countires should have in common globally and build a framework to make them standardized. The use of existing frameworks such as the Generation IV International Forum and the International Project on Innovative Nuclear Reactors and Fuel Cycles is considered effective to achieving this. Furthermore, a vigorous promotion such as international cooperative developments enables the formation of international consensus on major technologies for the fast reactor cycle as well as the saving of resources by infrastructure sharing. (author)

  1. 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)

  2. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident

    International Nuclear Information System (INIS)

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.)

  3. Nuclear Reactor RA Safety Report, Vol. 3, Building and installations

    International Nuclear Information System (INIS)

    RA reactor building is built of concrete and bricks as an enclosed building with limited number of controlled openings, and limited number of doors and windows. It is made of three parts: central; circular annex in the central part; sanitary corridor. The largest part of the RA reactor building is the reactor hall. This volume includes detailed description, figure and diagrams showing building characteristics, power supply systems, water supply systems, ventilation and heating systems, gas and compressed air installation as well as fire prevention system

  4. Magnetic divertors for experimental Tokamaks and fusion reactors

    International Nuclear Information System (INIS)

    Brief reports of working group discussions. These covered the requirements for a divertor in a fusion reactor including reducing impurities, exhausting the plasma and controlling the plasma-wall interactions. Divertor configurations were also reviewed and their merits and disadvantages compared. Existing divertor experiments were summarised and recommendations for further work made. Then the problems anticipated in designing a divertor for a conceptual reactor were considered. The physics of divertors and the scrape-off layer was discussed with reference to present models of plasma in divertors. Finally, experiments needed to demonstrate the feasibility of divertors for reactors and the development of specialised diagnostics for such experiments were considered. (U.K.)

  5. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Elimination, or substantial reduction, of the trade in highly enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less than 20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. A program is now beginning in the U.S. to develop the necessary fuel technology, but several years of work will be needed. Accordingly, as an immediate interim step, the U.S. is proposing to convert existing research and test reactors (and new designs) from the use of 90-93% enriched fuel to the use of 30-45% enriched fuel wherever this can be done without unacceptable reactor performance degradation

  6. Modern research reactors: design features and safety aspects

    International Nuclear Information System (INIS)

    The purpose of this article is to give a general information about the new orientations, which have been taken in the design and equipment of nuclear research reactors, and its wide uses in the area of basic and applied scientific research. these reactors have been subdivided into different categories according to their neutron flux density. In each category some physical and technical specifications were given for chosen examples. We end this article with a survey about the safety aspects related to its meaning in designing and operating of these reactors. (author). 5 refs., 4 figs

  7. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  8. Technology and use of low power research reactors

    International Nuclear Information System (INIS)

    The report contains a summary of discussions and 10 papers presented at the Consultants' Meeting on the Technology and Use of Low Power Research Reactors organized by the IAEA and held in Beijing (China) during 30 April - 3 May 1985. The following topics have been covered: reactor utilization in medicine and biology, in universities, for training, as a neutron source for radiography and some remarks on the safety of low power research reactors. A separate abstract was prepared for each paper presented at the meeting

  9. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  10. Novel reactors and energy synergetics status 1982

    International Nuclear Information System (INIS)

    The recession, increasing energy costs, recent studies like NASAP and INFCE, recent innovations and new developments have resulted in a new situation in the energy field. Even near term nuclear power R and D planning requires thus concurrent studies of spallation (accelerator) and fusion/fission hybrid breeding. A first overview of these and other novel reactors is presented. It is now realized more than before that the energy production must be based on optimal synergetics based on symbiotic systems that include a larger variety of energy sources, even if we restrict us, as in this report, to nuclear power. A central factor is the considerations associated with the constraints of fuel supplies, of enriched fissile fuels, of U and Th and of fusile fuels (T). This report emphasizes the inherent characteristics of various energy producing machines and symbiotic systems in this respect including the status, national programmes, environmental impacts and their expected break-even U-prices as reported in the literature. (Author)

  11. Survey of methods and measurements of nuclear reactor time and frequency responses

    International Nuclear Information System (INIS)

    Methods of measuring reactivity effects in nuclear reactors are described and the main control engineering analytical problems in nuclear reactors are detailed. A description of the use of reactor models and adaptive control in improving the economy of power producing nuclear reactors is included. (author)

  12. Decontamination and decommissioning of Shippingport commercial reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schreiber, J. [Dept. of Energy, Pittsburgh, PA (United States)

    1989-11-01

    To a certain degree, the decontamination and decommissioning (D and D) of the Shippingport reactor was a joint venture with Duquesne Light Company. The structures that were to be decommissioned were to be removed to at least three feet below grade. Since the land had been leased from Duquesne Light, there was an agreement with them to return the land to them in a radiologically safe condition. The total enclosure volume for the steam and nuclear containment systems was about 1.3 million cubic feet, more than 80% of which was below ground. Engineering plans for the project were started in July of 1980 and the final environmental impact statement (EIS) was published in May of 1982. The plant itself was shut down in October of 1982 for end-of-life testing and defueling. The engineering services portion of the decommissioning plans was completed in September of 1983. DOE moved onto the site and took over from the Navy in September of 1984. Actual physical decommissioning began after about a year of preparation and was completed about 44 months later in July of 1989. This paper describes the main parts of D and D.

  13. Health Physics Research Reactor (HPRR) operating experience and applications

    International Nuclear Information System (INIS)

    The Health Physics Research Reactor (HPRR) is a small, unmoderated fast pulse reactor located at the Oak Ridge National Laboratory (ORNL). The HPRR is the principle research tool of ORNL's Dosimetry Applications Research Group. The reactor is described, and its operating experience is presented. The HPRR is used by dosimeter vendors, government laboratories, nuclear power utilities, the military, and universities as well as by the ORNL staff for a wide variety of applications. These applications have been divided into six categories as follows: (1) biological effects studies, (2) criticality alarm testing, (3) dosimetry intercomparison studies, (4) neutron and gamma dose equivalent dosimeter development, (5) simulation of nuclear weapon spectra, and (6) training

  14. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  15. Sensitivity and Uncertainty Study for Thermal Molten Salt Reactors

    Science.gov (United States)

    Bidaud, Adrien; Ivanona, Tatiana; Mastrangelo, Victor; Kodeli, Ivo

    2006-04-01

    The Thermal Molten Salt Reactor (TMSR) using the thorium cycle can achieve the GEN IV objectives of economy, safety, non-proliferation and durability. Its low production of higher actinides, coupled with its breeding capabilities - even with a thermal spectrum - are very valuable characteristics for an innovative reactor. Furthermore, the thorium cycle is more flexible than the uranium cycle since only a small fissile inventory (reactor. The potential of these reactors is currently being extensively studied at the CNRS and EdF /1,2/. A simplified chemical reprocessing is envisaged compared to that used for the former Molten Salt Breeder Reactor (MSBR). The MSBR concept was developed at Oak Ridge National Laboratory (ORNL) in the 1970's based on the Molten Salt Reactor Experiment (MSRE). The main goals of our current studies are to achieve a reactor concept that enables breeding, improved safety and having chemical reprocessing needs reduced and simplified as much as reasonably possible. The neutronic properties of the new TMSR concept are presented in this paper. As the temperature coefficient is close to zero, we will see that the moderation ratio cannot be chosen to simultaneously achieve a high breeding ratio, long graphite lifetime and low uranium inventory. It is clear that any safety margin taken due to uncertainty in the nuclear data will significantly reduce the capability of this concept, thus a sensitivity analysis is vital to propose measurements which would allow to reduce at present high uncertainties in the design parameters of this reactor. Two methodologies, one based on OECD/NEA deterministic codes and one on IPPE (Obninsk) stochastic code, are compared for keff sensitivity analysis. The uncertainty analysis of keff using covariance matrices available in evaluated files has been performed. Furthermore, a comparison of temperature coefficient sensitivity profiles is presented for the most important reactions. These results are used to review the

  16. Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hho jung; Seul, K.W.; Ahn, S.K.; Bang, Y.S.; Park, D.G.; Kim, B.K.; Kim, W.S.; Lee, J.H.; Kim, W.K.; Shim, T.M.; Choi, H.S.; Ahn, H.J.; Jung, D.W.; Kim, G.I.; Park, Y.M.; Lee, Y.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1997-07-01

    The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

  17. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  18. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  19. Comparison of simulated and measured quantities of a duplex reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koskela, M.; Kajava, M. [ABB Marine, Helsinki (Finland)

    1997-12-31

    The purpose of this article is to illustrate the use of an analog simulator as a design tool when designing new power electric equipment. The purpose of simulation is to predict the functionality of electrical equipment to be constructed. Duplex reactor is an electromagnetic device designed to reduce voltage harmonics and short circuit currents in the ship electrical network. In this report a comparison between simulated and measured electrical quantities of a duplex reactor has been made. The purpose of the measurements was to show the correct functioning of the reactor. The simulation results and the measured waveforms corresponds well to each other. (orig.) 4 refs.

  20. Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976

    International Nuclear Information System (INIS)

    During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation

  1. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems: Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well establish

  2. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  3. Increasing Fuel Utilization of Breed and Burn Reactors

    Science.gov (United States)

    Di Sanzo, Christian Diego

    Breed and Burn reactors (B&B), also referred to Traveling Wave Reactors, are fast spectrum reactors that can be fed indefinitely with depleted uranium only, once criticality is achieved without the need for fuel reprocessing. Radiation damage to the fuel cladding limits the fuel utilization of B&B reactors to ˜ 18-20% FIMA (Fissions of Initial Metal Atoms) -- the minimum burnup required for sustaining the B&B mode of operation. The fuel discharged from this type of cores contain ˜ 10% fissile plutonium. Such a high plutonium content poses environmental and proliferation concerns, but makes it possible to utilize the fuel for further energy production. The objectives of the research reported in this dissertation are to analyze the fuel cycle of B&B reactors and study new strategies to extend the fuel utilization beyond ˜ 18-20% FIMA. First, the B&B reactor physics is examined while recycling the fuel every 20% FIMA via a limited separation processing, using either the melt refining or AIROX dry processes. It was found that the maximum attainable burnup varies from 54% to 58% FIMA -- depending on the recycling process and on the fraction of neutrons lost via leakage and reactivity control. In Chapter 3 the discharge fuel characteristics of B&B reactors operating at 20% FIMA and 55% FIMA is analyzed and compared. It is found that the 20% FIMA reactor discharges a fuel with about ˜ 80% fissile plutonium over total plutonium content. Subsequently a new strategy of minimal reconditioning, called double cladding is proposed to extend the fuel utilization in specifically designed second-tier reactors. It is found that with this strategy it is possible to increase fuel utilization to 30% in a sodium fast reactor and up to 40% when a subcritical B&B core is driven by an accelerator-driven spallation neutron source. Lastly, a fuel cycle using Pressurized Water Reactors (PWR) to reduce the plutonium content of discharged B&B reactors is analyzed. It was found that it is

  4. Isotopic composition and neutronics of the Okelobondo natural reactor

    Science.gov (United States)

    Palenik, Christopher Samuel

    The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve

  5. Modelling and control design for SHARON/Anammox reactor sequence

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation and for optimal operation. Work in progress on quantitatively analysing different control structure......With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...

  6. Accelerators and nuclear reactors as tools in hot atom chemistry

    International Nuclear Information System (INIS)

    The characteristics of accelerators and of nuclear reactors - the latter to a lesser extent - are discussed in view of their present and future use in hot atom chemistry research and its applications. (author)

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

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

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

  8. Economic viability of innovative nuclear reactor and fuel cycle technologies

    International Nuclear Information System (INIS)

    Full text: Nuclear power has established its position as one of the most stable electricity supply sources in many countries in the world, supplying about 17% of total electricity generated. However, in order to keep that position, there are two important challenges that nuclear energy will face in the coming decades. They are: competition, and social/political acceptance (including non-proliferation and terrorism). There is an increasing concern that existing nuclear technologies may not be able to overcome such tough challenges. It is expected that innovative technologies can be a part of the solutions to overcome such challenges. This paper focuses on economic viability of innovative nuclear reactor and its associated fuel cycle technologies. First, it is important to consider the long term energy paths and potential role of nuclear power under different scenarios. We applied global energy optimization model based on IPCC scenarios. Then, we look at Japan, where electricity market is being liberalized, in order to explore how liberalization will have influence economic viability of nuclear power. The following are our basic conclusions: CO2 constraints as well as power generation cost competitiveness could affect future growth of nuclear power quite significantly. Current trend suggests that nuclear power would not grow much without CO2 constraints, or even face minus growth if its power generation cost became higher. On the other hand, cost reduction with CO2 constraints could accelerate future expansion of nuclear power quite significantly; In addition to life-long average generation cost, other investment criteria (such as asset productivity) may become critically important under the liberalized market. Under the liberalized electricity market, short term investment criteria could become more important than 30 year life time average cost. This suggests that small initial investment is more acceptable than large capital investment. Advanced nuclear reactor

  9. Considerations and Infrastructure Milestones for a Research Reactor Project

    International Nuclear Information System (INIS)

    Establishment of a research reactor is a major project requiring careful planning, preparation, implementation, and investment in time and human resources. The implementation of such a project requires establishment of sustainable infrastructures, including legal and regulatory, safety, technical, and economic. This paper discusses the scope of these infrastructures and the major stages in their development; starting with a robust pre-project justification for the research reactor and moving through three milestones in the establishment of the infrastructure itself. The paper discusses also the main elements of the feasibility study for a new research reactor project and specific safety and technical considerations in different phases of the project as well as the major activities to be performed along with the project phases, including progressive involvement of the main organizations in the project, and application of the IAEA Code of Conduct on the Safety of Research Reactors and IAEA Safety Standards. (author)

  10. Modeling of Reactor Kinetics and Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Johnson; Scott Lucas; Pavel Tsvetkov

    2010-09-01

    In order to model a full fuel cycle in a nuclear reactor, it is necessary to simulate the short time-scale kinetic behavior of the reactor as well as the long time-scale dynamics that occur with fuel burnup. The former is modeled using the point kinetics equations, while the latter is modeled by coupling fuel burnup equations with the kinetics equations. When the equations are solved simultaneously with a nonlinear equation solver, the end result is a code with the unique capability of modeling transients at any time during a fuel cycle.

  11. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  12. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  13. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  14. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. PMID:27108375

  15. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Rosen, Lee J.; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-09-27

    A commercially viable modular ceramic oxygen transport membrane reforming reactor for producing a synthesis gas that improves the thermal coupling of reactively-driven oxygen transport membrane tubes and catalyst reforming tubes required to efficiently and effectively produce synthesis gas.

  16. Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

    International Nuclear Information System (INIS)

    Producing nuclear energy in order to reduce the anthropic CO2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

  17. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983

    International Nuclear Information System (INIS)

    After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW

  18. Advances in sodium technology, testing and diagnostics of fast reactors

    International Nuclear Information System (INIS)

    The collection contains a selection of 29 papers from three international specialists' meetings: the CMEA conference ''Control and measuring instruments and diagnostic systems of fast reactors'' held in the GDR in April 1983; the IAEA conference on nuclear power experience held in Austria in September 1982; and the conference ''Problems of technology and corrosion in sodium coolant and protective gas'' held in the GDR in April 1977. Three papers on operating experience with Soviet fast reactors and their safety have a general character; they are followed up by three papers on sodium technology. Five papers deal with the diagnostics of fast sodium cooled reactors and nine papers are devoted to the diagnostics of steam generators. Eight papers relate to detectors for the diagnostics of fast reactors. Safety regulations for work with alkali metals are added. (A.K.)

  19. Beta limits in EBT and their implications for a reactor

    International Nuclear Information System (INIS)

    Because of uncertainties in extrapolating results of simplified models to a reactor plasma, the parameters that influence the beta limits cannot be determined accurately at the present time. Also, reasonable changes within the models and/or assumptions are seen to affect the core beta limits by almost an order of magnitude. Hence, at the present, these limits cannot be used as rigid (and reliable) requirement for ELMO Bumpy Torus (EBT) reactor engineering considerations. However, sensitivity studies can be carried out to determine the boundaries of the operating regime and to demonstrate the effects of various modes, assumptions, and models on reactor performance (Q value). First the modes believed to limit the core β and ring plasma performance are discussed, and the simplifications and/or assumptions involved in deriving these limits are highlighted. Then, the implications of these limits for a reactor are given

  20. Research reactor decommissioning experience - concrete removal and disposal -

    International Nuclear Information System (INIS)

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

  1. Design and development of small and medium integral reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR`s, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs.

  2. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

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

  3. Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis II. Quantification of inhibition and suitability of membrane reactors

    DEFF Research Database (Denmark)

    Andric, Pavle; Meyer, Anne S.; Jensen, Peter Arendt;

    2010-01-01

    ideal reactor types, i.e. batch, continuous stirred, and plug-flow, is illustrated quantitatively by modeling different extents of cellulose conversion at different reaction conditions. The main operational challenges of membrane reactors for lignocellulose conversion are highlighted. Key membrane...... of the available literature data for glucose removal by membranes and for cellulose enzyme stability in membrane reactors are given. The treatise clearly shows that membrane reactors allowing continuous, complete, glucose removal during enzymatic cellulose hydrolysis, can provide for both higher cellulose...... hydrolysis rates and higher enzyme usage efficiency (kg(product/)kg(enzyme)). Current membrane reactor designs are however not feasible for large scale operations. The report emphasizes that the industrial realization of cellulosic ethanol requires more focus on the operational feasibility within...

  4. Some Movement Mechanisms and Characteristics in Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available The pebblebed-type high temperature gas-cooled reactor is considered to be one of the promising solutions for generation IV advanced reactors, and the two-region arranged reactor core can enhance its advantages by flattening neutron flux. However, this application is held back by the existence of mixing zone between central and peripheral regions, which results from pebbles’ dispersion motions. In this study, experiments have been carried out to study the dispersion phenomenon, and the variation of dispersion region and radial distribution of pebbles in the specifically shaped flow field are shown. Most importantly, the standard deviation of pebbles’ radial positions in dispersion region, as a quantitative index to describe the size of dispersion region, is gotten through statistical analysis. Besides, discrete element method has been utilized to analyze the parameter influence on dispersion region, and this practice offers some strategies to eliminate or reduce mixing zone in practical reactors.

  5. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  6. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

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

  7. Studies on the transient operation and stability of fast reactors

    International Nuclear Information System (INIS)

    These studies form part of the general programme of perfecting calculation methods for fast reactors. The basic formulae are given for the layouts used, i.e. the classic kinetic and thermal exchange equations, etc. A description is then given of the digital computer methods employed for studying the stable functioning of the reactor and of the methods used for transient operation studies. Finally, some examples of application are discussed and a comparison is made with parallel studies on the same subject. (author)

  8. The need and prospects for improved fusion reactors

    Science.gov (United States)

    Krakowski, R. A.; Miller, R. L.; Hagenson, R. L.

    1986-09-01

    Conceptual fusion reactor studies over the past 10-15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100-200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.

  9. Nuclear Data and the Oklo Natural Nuclear Reactors

    Science.gov (United States)

    Gould, C. R.; Sharapov, E. I.; Sonzogni, A. A.

    2014-04-01

    Data from the Oklo natural nuclear reactors have enabled some of the most sensitive terrestrial tests of time variation of dimensionless fundamental constants. The constraints on variation of αEM, the fine structure constant are particular good, but depend on the reliability of the nuclear data, and on the reliability of the modeling of the reactor environment. We briefly review the history of these tests and discuss our recent work in 1) attempting to better bound the temperatures at which the reactors operated, 2) investigating whether the γ-ray fluxes in the reactors could have contributed to changing lutetium isotopic abundances and 3) determining whether lanthanum isotopic data could provide an alternate estimate of the neutron fluence.

  10. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  11. Development of Core Design Model for Small-Sized Research Reactor and Establishment of Infrastructure for Reactor Export

    International Nuclear Information System (INIS)

    Within 10 years a growing world-wide demand of new research reactor construction is expected because of obsolescence. In Korea, a new research reactor is also required in order to meet domestic demand of utilization. KAERI has been devoted to develop an export-oriented research reactors for these kinds of demand. A next generation research reactor should comply with general requirements for safety, economics, environment-friendliness and non-proliferation as well as high performance requirement of high flux level. A export-tailored reactor should be developed for the demand of developing counties or under-developed countries. A new design concept is to be developed for a long cycle length core which has excellent irradiation facility with high flux

  12. Recent experience and new developments in reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    In this paper, Framatome's recent experience and new developments in the manufacture of pressurized water reactor (PWR) reactor pressure vessel (RPV)s is described to show the very high standards of quality achieved to meet the most stringent requirements. Outstanding new developments include: qualification and utilisation of thick forged shell rings made from large hollow ingots; fully automatic submerged arc narrow gap welding; electroslag stainless steel cladding process; nozzle buttering by automatic hotwire TIG process

  13. Cold nuclear fusion reactor and nuclear fusion rocket

    OpenAIRE

    Huang Zhenqiang

    2013-01-01

    "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nucle...

  14. Study and Analysis on Naphtha Catalytic Reforming Reactor Simulation

    Institute of Scientific and Technical Information of China (English)

    Liang Ke min; Song Yongji; Pan Shiwei

    2004-01-01

    A naphtha catalytic reforming unit with four reactors connected in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reaction characteristics based on idealizing the complex naphtha mixture to represent the paraffin, naphthene, and aromatic groups with individual compounds. The simulation results based on above models agree very well with actual operating data of process unit.

  15. A study on naphtha catalytic reforming reactor simulation and analysis

    Institute of Scientific and Technical Information of China (English)

    LIANG Ke-min; GUO Hai-yan; PAN Shi-wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  16. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  17. Nuclear data and reactor physics activities in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

    1998-03-01

    The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

  18. High burnup fast reactor fuel: processing and waste management experiences

    International Nuclear Information System (INIS)

    The routine processing of mixed Plutonium/Uranium oxide fuels from the Prototype Fast Reactor (PFR) at Dounreay began in September 1980 and the design features of the modified Dounreay Fast Reactor (DFR) reprocessing plant and experience of the first active campaign were described in a paper to the British Nuclear Engineering Society in November 1981 (1). Since then progress in processing the fuel discharged from PFR has been covered briefly in a number of papers to international conferences and the Public Inquiry held in 1986 into the outline planning application for the proposed European Demonstration Reprocessing Plant. During this decade considerable experience in the operation of fast reactors and associated fuel plants has been accumulated providing confidence in the system before entering the next development phase - that of its commercial demonstration. Confidence in the UK draws on the successful operation of the PFR and the associated Dounreay fuel reprocessing and BNF Sellafield fabrication plants. Of equal importance is public confidence in safe operation and in the management of wastes generated by a fast reactor system. The present paper is a review of fast reactor reprocessing and waste management at the Dounreay Nuclear Establishment (DNE) as a contribution to the present status of the fast reactor system

  19. Safety analysis for small and medium size integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jin H.; Chang, M. H.; Bae, K. H.; Lim, H. S.; Kwon, M.; Lee, Y. J.; Hwang, Y. D.; Kim, S. O.; Lee, W. J.; Chung, B. D.

    1997-09-01

    Sets of safety and performance related design basis events have been proposed for the SMART. Detailed descriptions of the events, justification and the selection criteria are specified. Operation modes of the SMART integral reactor are described. Safety systems as well as the components specific to the SMART integral reactor are evaluated. Thermal hydraulic system codes are evaluated for the use of the safety and performance analysis. Both the safety and performance methodology as well as the code systems are proposed for the safety and performance analysis of the SMART integral reactor. A preliminary PIRT for the SMART integral reactor was developed by an expert panel during the study. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect test was developed for the thermal hydraulic model development and the system code validation. This experimental program will be also used to evaluated the safety systems and to support licensing confirmation of the SMART integral reactor. The results of the study will be used for the conceptual design of the SMART integral reactor. (author). 58 refs., 23 tabs., 30 figs.

  20. Proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic

    International Nuclear Information System (INIS)

    The proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic - 6. ENFIR - allow to evaluate the present status of development in reactor physics and thermohydraulic fields. The mathematical models and methods for calculating neutronic of nuclear reactors, safety reactor analysis, measuring methods of neutronic parameters, computerized simulation of accidents, transients and thermohydraulic analysis are presented. (M.C.K.)

  1. Reactor technology assessment and selection utilizing systems engineering approach

    Science.gov (United States)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  2. Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor

    Science.gov (United States)

    Bingham, Dennis N.; Klingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.

    2012-11-13

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  3. The DALAT nuclear research reactor operation and conversion status

    International Nuclear Information System (INIS)

    This paper presents operation and conversion status of the DALAT Nuclear Research Reactor (DNRR). The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA-MARK II reactor. The core is loaded with Soviet-designed standard type WWR-M2 fuel assemblies with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. The DNRR is operated mainly in continuous runs of 100 hours, once every 4 weeks, for radioisotope production, neutron activation analyses, training and research purposes. The remaining time between two continuous runs, is devoted to maintenance activities and to short runs. Until now 4 fuel reloading were executed. The reactor control and instrumentation system was upgraded in 1994. And now the reactor control system is being replaced by new one, the replacement will be fulfilled in March 2007. The study on fuel conversion has been done on the basis of a new LEU of 19.75% with UO2-Al dispersion fuel meat instead of the current HEU of 36% with aluminium-uranium alloy. The results of the study show that operation time of mixed core by inserting 36 LEU fuel assemblies lasts much longer than by inserting 36 HEU fuel assemblies (14.5 instead of 10.5 years). Neutron flux performances at irradiation positions are not significantly changed. Now we are working for realizing fuel conversion of the DNRR

  4. R- AND P- REACTOR VESSEL IN-SITU DECOMISSIONING VISUALIZATION

    Energy Technology Data Exchange (ETDEWEB)

    Vrettos, N.; Bobbitt, J.; Howard, M.

    2010-06-07

    The R- & P- Reactor facilities were constructed in the early 1950's in response to Cold War efforts. The mission of the facilities was to produce materials for use in the nation's nuclear weapons stockpile. R-Reactor was removed from service in 1964 when President Johnson announced a slowdown of he nuclear arms race. PReactor continued operation until 1988 until the facility was taken off-line to modernize the facility with new safeguards. Efforts to restart the reactor ended in 1990 at the end of the Cold War. Both facilities have sat idle since their closure and have been identified as the first two reactors for closure at SRS.

  5. Optimized Design and Discussion on Middle and Large CANDLE Reactors

    Directory of Open Access Journals (Sweden)

    Xiaoming Chai

    2012-08-01

    Full Text Available CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scale fission nuclear energy utilization for thousands of years, benefitting the whole of society. For many developing countries with a large population and high energy demands, such as China and India, middle (1000 MWth and large (2000 MWth CANDLE fast reactors are obviously more suitable than small reactors [2]. In this paper, the middle and large CANDLE reactors are investigated with U-Pu and combined ThU-UPu fuel cycles, aiming to utilize the abundant thorium resources and optimize the radial power distribution. To achieve these design purposes, the present designs were utilized, simply dividing the core into two fuel regions in the radial direction. The less active fuel, such as thorium or natural uranium, was loaded in the inner core region and the fuel with low-level enrichment, e.g. 2.0% enriched uranium, was loaded in the outer core region. By this simple core configuration and fuel setting, rather than using a complicated method, we can obtain the desired middle and large CANDLE fast cores with reasonable core geometry and thermal hydraulic parameters that perform safely and economically; as is to be expected from CANDLE. To assist in understanding the CANDLE reactor’s attributes, analysis and discussion of the calculation results achieved are provided.

  6. Knowledge management in fast reactors and related fuel cycles

    International Nuclear Information System (INIS)

    Full text: The 21st century is ushering in a new phase of economic and social development which can be referred as 'Knowledge Economy', in which knowledge has become the key asset in determining the organization's success or failure. The IAEA defines knowledge management as an integrated, systematic approach to identify, manage and share an organization's knowledge collectively in order to help achieve the objectives of the organization. Nuclear technology is very complex and a highly technical endeavor. It relies on innovative creation, storage and dissemination of knowledge. The nuclear energy is characterized by long time scales and technological excellence. Nuclear knowledge management is a critical input to nuclear power industry, the associated fuel cycle activities and nuclear applications in medicine, industry and agriculture. Realizing the importance of knowledge preservation in the area of fast reactor technology, IAEA had given a consultancy work to Argonne National Laboratory to study and suggest the means of knowledge management. The IAEA initiative seeks to establish a comprehensive inventory of fast reactor data and knowledge for the fast reactor development in the coming years. It was suggested that the knowledge regarding important disciplines like fuels and materials, reactor physics and core design, operations, the demonstration of safety should be preserved. Various countries have initiated the fast reactor knowledge preservation activities. In France, CEA, EDF and Framatome ANP have initiated liquid metal cooled fast reactor knowledge preservation project that deals with R and D aspects and Superphenix design. European Fast Reactor collaboration (MASURCA,SNEAK,ZEBRA) has preserved the zero power critical experimental data in the SNEDAX database. Japan has started a comprehensive knowledge preservation program including the capture of 'Human Knowledge' based on interviews. In Russia steps are initiated to preserve fast reactor knowledge

  7. Proceedings of the topical meeting on reactor physics and safety: Sessions 1-10. Volume 1

    International Nuclear Information System (INIS)

    Technical papers and invited lectures presented at the International Topical Meeting on Reactor Physics and Safety are presented. The sessions include a general session on Challenges in Reactor Physics and Safety. Together with sessions on conventional reactor physics topics, there are sessions on safety in nuclear design, dynamic behavior of reactors, degraded cores, research reactors and pressure vessel embrittlement. This conference is broad in scope and brings together experts from all over the fee world to present papers and exchange ideas on the reactor physics and safety aspects of nuclear reactors

  8. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  9. Fiber-Optical Sensors: Basics and Applications in Multiphase Reactors

    Directory of Open Access Journals (Sweden)

    Guozheng Li

    2012-09-01

    Full Text Available This work presents a brief introduction on the basics of fiber-optical sensors and an overview focused on the applications to measurements in multiphase reactors. The most commonly principle utilized is laser back scattering, which is also the foundation for almost all current probes used in multiphase reactors. The fiber-optical probe techniques in two-phase reactors are more developed than those in three-phase reactors. There are many studies on the measurement of gas holdup using fiber-optical probes in three-phase fluidized beds, but negative interference of particles on probe function was less studied. The interactions between solids and probe tips were less studied because glass beads etc. were always used as the solid phase. The vision probes may be the most promising for simultaneous measurements of gas dispersion and solids suspension in three-phase reactors. Thus, the following techniques of the fiber-optical probes in multiphase reactors should be developed further: (1 online measuring techniques under nearly industrial operating conditions; (2 corresponding signal data processing techniques; (3 joint application with other measuring techniques.

  10. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  11. Application of non-destructive testing and in-service inspections to research reactors and preparation of ISI programme and manual for WWR-C research reactors

    International Nuclear Information System (INIS)

    The present report gives a review on the results of application of non-destructive testing and in-service inspections to WWR-C reactors in different countries. The major problems related to reactor safety and the procedure of inspection techniques are investigated to collect the experience gained from this type of reactors. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of their rehabilitation programmes. 9 figs., 4 tabs

  12. Analysis of sodium-cooled fast reactor operations world-wide and consequences for future reactor design and operation

    International Nuclear Information System (INIS)

    In 2007, the sodium-cooled fast reactors which are currently operating or have operated throughout the world will have accumulated collective experience of 379 years of operation for 18 different reactors. This paper summarizes the various incidents and problems which have impacted these reactors' operations, ranks them by function and lists the solutions which were brought. This paper solely looks at the operating problems which occurred in the 'sodium' part of the fast reactors. Problems encountered in the classic water/steam part are not described herein review is made for all these reactors on: The water/sodium reactions occurred in the steam generators; -The technical difficulties on the primary components (pumps and exchangers) and the repair operations; -The incidents in handling operations; - The spurious leaks or transfers of sodium and their consequences; - The intakes of air or impurities; -The experience from fuel and clad failures; - The neutronic operations and control; -The material behaviour problems (as 321 or SPX1 drum); - The difficulties due to sodium aerosols. And all the specific difficulties due to sodium related technological problems. This paper also describes how the experience gained has been taken into account in the safe operation of these plants and also in the design of future reactors. n conclusion it appears that a significant experience has been accumulated that allow today good availability for the remaining operating plants and also good design possibilities for the future. (author)

  13. Reactor pressure vessel head vents and methods of using the same

    Science.gov (United States)

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  14. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  15. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  16. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  17. Design of a Compact and Versatile Bench Scale Tubular Reactor

    Directory of Open Access Journals (Sweden)

    R. Prasad

    2009-06-01

    Full Text Available A compact and versatile laboratory tubular reactor has been designed and fabricated keeping in view of reducing capital cost and minimising energy consumption for gas/vapor-phase heterogeneous catalytic reactions. The reactor is consisted of two coaxial corning glass tubes with a helical coil of glass tube in between the coaxial tubes serving as vaporiser and pre-heater, the catalyst bed is in the inner tube. A schematic diagram of the reactor with detailed dimensions and working principles are described. The attractive feature of the reactor is that the vaporiser, pre-heater and fixed bed reactor are merged in a single compact unit. Thus, the unit minimises separate vaporiser and pre-heater, also avoids separate furnaces used for them and eliminate auxiliary instrumentation such as temperature controller etc. To demonstrate the system operation and illustrate the key features, catalyst screening data and the efficient collection of complete, and accurate intrinsic kinetic data are provided for oxidation of CO over copper chromite catalyst. CO oxidation is an important reaction for auto-exhaust pollution control. The suitability of the versatile nature of the reactor has been ascertained for catalytic reactions where either volatile or vaporizable feeds can be introduced to the reaction zone, e.g. oxidation of iso-octane, reduction of nitric oxide, dehydrogenation of methanol, ethanol and iso-propanol, hydrogenation of nitrobenzene to aniline, etc. Copyright (c 2009 by BCREC. All Rights reserved.[Received: 10 February 2009, Accepted: 9 May 2009][How to Cite: R. Prasad, G. Rattan. (2009. Design of a Compact and Versatile Bench Scale Tubular Reactor. Bulletin of Chemical Reaction Engineering and Catalysis, 4(1: 5-9.  doi:10.9767/bcrec.4.1.1250.5-9][How to Link/ DOI: http://dx.doi.org/10.9767/bcrec.4.1.1250.5-9

  18. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  19. Design review and hazop studies for stable salt reactor

    International Nuclear Information System (INIS)

    Atkins has been assisting Moltex Energy in carrying out a review of their Stable Salt Reactor (SSR) (formerly the Simple Molten Salt Reactor (SMSR)), which they believe has many benefits over other nuclear reactor systems. Their current design is for a 2.5 GW thermal output reactor, supplying a superheated steam turbine that could provide an electrical output of 1GW. Our first task was to carry out an assessment of the SSR concept against relevant UK Regulatory Requirements, using data and information that was provided by Moltex Energy. Moltex Energy updated their design using the output from this review. The next activity was to carry out a Hazard and Operability (HAZOP) review to identify issues that may not have been apparent in the original design and the UK Regulatory Requirements assessment. For the HAZOP, we brought together a Process Engineer, Fuel Route and Mechanical Handling Specialist, Safety Case Engineer, Reactor Chemist, Decommissioning and Waste Management Specialist. The HAZOP review gave rise to the generation of a list of key Structures, Systems and Components (SSCs) that would be necessary in a fully designed reactor system, which were discussed and described in a report. This report provided the scope and assumptions that were used as the basis for the costing estimate, using expertise provided in-house by Faithful and Gould. Our paper will discuss the processes used in more detail, identify how these processes increased the knowledge and design concepts to ascertain the SSCs in more detail. Discuss the costing approach and the use of a three-point capital cost estimate, modelled using Monte Carlo simulation, to provide a cost/uncertainty distribution profile. Discussion on how Atkins can assist with the Indian reactor programme using a similar approach detailed above

  20. System and method for air temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  1. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    Science.gov (United States)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  2. International Seminar on Gasification 2009 - Biomass Gasification, Gas Clean-up and Gas Treatment

    Energy Technology Data Exchange (ETDEWEB)

    2009-10-15

    During the seminar international and national experts gave presentations concerning Biomass gasification, Gas cleaning and gas treatment; and Strategy and policy issues. The presentations give an overview of the current status and what to be expected in terms of development, industrial interest and commercialization of different biomass gasification routes. The following PPT presentations are reproduced in the report: Black Liquor Gasification (Chemrec AB.); Gasification and Alternative Feedstocks for the Production of Synfuels and 2nd Generation Biofuels (Lurgi GmbH); Commercial Scale BtL Production on the Verge of Becoming Reality (Choren Industries GmbH.); Up-draft Biomass Gasification (Babcock and Wilcox Voelund A/S); Heterogeneous Biomass Residues and the Catalytic Synthesis of Alcohols (Enerkem); Status of the GoBiGas-project (Goeteborg Energi AB.); On-going Gasification Activities in Spain (University of Zaragoza,); Biomass Gasification Research in Italy (University of Perugia.); RDandD Needs and Recommendations for the Commercialization of High-efficient Bio-SNG (Energy Research Centre of the Netherlands.); Cleaning and Usage of Product Gas from Biomass Steam Gasification (Vienna University of Technology); Biomass Gasification and Catalytic Tar Cracking Process Development (Research Triangle Institute); Syngas Cleaning with Catalytic Tar Reforming (Franhofer UMSICHT); Biomass Gas Cleaning and Utilization - The Topsoee Perspective (Haldor Topsoee A/S); OLGA Tar Removal Technology (Dahlman); Bio-SNG - Strategy and Activities within E.ON (E.ON Ruhrgas AG); Strategy and Gasification Activities within Sweden (Swedish Energy Agency); 20 TWh/year Biomethane (Swedish Gas Association)

  3. EMAT weld inspection and weld machine diagnostic system for continuous coil processing lines

    Science.gov (United States)

    Latham, Wayne M.; MacLauchlan, Daniel T.; Geier, Dan P.; Lang, Dennis D.

    1996-11-01

    Weld breaks of steel coil during cold rolling and continuous pickling operations are a significant source of lost productivity and product yield. Babcock and Wilcox Innerspec Technologies has developed a weld process control system which monitors the key variables of the welding process and determines the quality of the welds generated by flash butt welding equipment. This system is known as the Temate 2000 Automated Flash Butt Weld Inspection and Weld Machine Diagnostic System. The Temate 2000 system utilizes electro- magnetic acoustic transducer (EMAT) technology as the basis for performing on-line, real-time, nondestructive weld quality evaluation. This technique accurately detects voids, laps, misalignment and over/under trim conditions in the weld. Results of the EMAT weld inspection are immediately presented to the weld machine operator for disposition. Welding process variables such as voltage, current, platen movements and upset pressures are monitored and collected with the high speed data acquisition system. This data is processed and presented in real-time display to indicate useful welding process information such as platen crabbing, upset force, peak upset current, and many others. Alarming for each variable is provided and allows detailed maintenance reports and summary information to be generated. All weld quality and process parameter data are stored, traceable to each unique weld, and available for post process evaluation. Installation of the Temate 2000 system in a major flat rolled steel mill has contributed to near elimination of weld breakage and increased productivity at this facility.

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

    International Nuclear Information System (INIS)

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm2.s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm2.s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U02-12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at 30 k

  5. Practices for Neutronic Design of Research Reactors: Safety and Performances

    International Nuclear Information System (INIS)

    In brief, the design aims to have a facility which is quickly operational and profitable, safe and able to evolve over 40 or 60 years, taking into account both the evolution of the requirements for experiments or production yet to be realized and the safety practices. This paper presents the AREVA current design and safety practices (both cannot be realized without the other) for the neutronic design of the research reactor (RR) cores. It completes the paper and presents the general methodology of neutronic design studies for the safety and performance aspects and only slightly focuses on the reactivity shutdown systems and the neutronic calculation schemes. The main points are illustrated with examples of the Jules Horowitz Reactor (core designer point of view). On this basis of our general methodology, certain problems are separated in order to permit rapid reiteration at an individual level before the final synthesis. For example: to carry out generic studies of fuel management strategies and core reactivity control in order to manage the power peak (need core depletion calculation) and to be able to reason step 0 for certain optimizations of the core geometry and characteristics. For the neutronic calculation scheme, our current practice is to combine the use of the deterministic and stochastic codes. The strong points of each type of code are used to reinforce the safety and the performance of our cores. In this field, AREVA has a R and D framework involving and coordinating the participants from the various sectors (power reactors, research reactor etc) in the development of the general calculation methods and associated tools, in particular for Monte Carlo core depletion calculations. The CEA (along with APOLLO, CRONOS and TRIPOLI codes) largely supports us in this field. Comparisons between MCNP and TRIPOLI and between the various libraries (ENDF, JEF, etc.) are also performed. That includes the recalculation of existing reactors (OSIRIS, ORPHEE, AZUR

  6. Tank Vapor Sampling and Analysis Data Package for Tank 241-Z-361 Sampled 09/22/1999 and 09/271999 During Sludge Core Removal

    Energy Technology Data Exchange (ETDEWEB)

    VISWANATH, R.S.

    1999-12-29

    This data package presents sampling data and analytical results from the September 22 and 27, 1999, headspace vapor sampling of Hanford Site Tank 241-2-361 during sludge core removal. The Lockheed Martin Hanford Corporation (LMHC) sampling team collected the samples and Waste Management Laboratory (WML) analyzed the samples in accordance with the requirements specified in the 241-2361 Sludge Characterization Sampling and Analysis Plan, (SAP), HNF-4371, Rev. 1, (Babcock and Wilcox Hanford Corporation, 1999). Six SUMMA{trademark} canister samples were collected on each day (1 ambient field blank and 5 tank vapor samples collected when each core segment was removed). The samples were radiologically released on September 28 and October 4, 1999, and received at the laboratory on September 29 and October 6, 1999. Target analytes were not detected at concentrations greater than their notification limits as specified in the SAP. Analytical results for the target analytes and tentatively identified compounds (TICs) are presented in Section 2.2.2 starting on page 2B-7. Three compounds identified for analysis in the SAP were analyzed as TICs. The discussion of this modification is presented in Section 2.2.1.2.

  7. Tank Vapor Sampling and Analysis Data Package for Tank 241-Z-361 Sampled 09/22/1999 and 09/27/1999, During Sludge Core Removal

    International Nuclear Information System (INIS)

    This data package presents sampling data and analytical results from the September 22 and 27, 1999, headspace vapor sampling of Hanford Site Tank 241-2-361 during sludge core removal. The Lockheed Martin Hanford Corporation (LMHC) sampling team collected the samples and Waste Management Laboratory (WML) analyzed the samples in accordance with the requirements specified in the 241-2361 Sludge Characterization Sampling and Analysis Plan, (SAP), HNF-4371/Rev. 1, (Babcock and Wilcox Hanford Corporation, 1999). Six SUMMA(trademark) canister samples were collected on each day (1 ambient field blank and 5 tank vapor samples collected when each core segment was removed). The samples were radiologically released on September 28 and October 4, 1999, and received at the laboratory on September 29 and October 6, 1999. Target analytes were not detected at concentrations greater than their notification limits as specified in the SAP. Analytical results for the target analytes and tentatively identified compounds (TICs) are presented in Section 2.2.2 starting on page 2B-7. Three compounds identified for analysis in the SAP were analyzed as TICs. The discussion of this modification is presented in Section 2.2.1.2

  8. Optimizing Reactors Selection and Sequencing:Minimum Cost versus Minimum Volume

    Institute of Scientific and Technical Information of China (English)

    Rachid Chebbi

    2014-01-01

    The present investigation targets minimum cost of reactors in series for the case of one single chemical reaction, considering plug flow and stirred tank reactor(s) in the sequence of flow reactors. Using Guthrie’s cost correlations three typical cases were considered based on the profile of the reaction rate reciprocal versus conversion. Significant differences were found compared to the classical approach targeting minimum total reactor volume.

  9. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

  10. Microalgal reactors: a review of enclosed system designs and performances.

    Science.gov (United States)

    Carvalho, Ana P; Meireles, Luís A; Malcata, F Xavier

    2006-01-01

    One major challenge to industrial microalgal culturing is to devise and develop technical apparata, cultivation procedures and algal strains susceptible of undergoing substantial increases in efficiency of use of solar energy and carbon dioxide. Despite several research efforts developed to date, there is no such thing as "the best reactor system"- defined, in an absolute fashion, as the one able to achieve maximum productivity with minimum operation costs, irrespective of the biological and chemical system at stake. In fact, choice of the most suitable system is situation-dependent, as both the species of alga available and the final purpose intended will play a role. The need of accurate control impairs use of open-system configurations, so current investigation has focused mostly on closed systems. In this review, several types of closed bioreactors described in the technical literature as able to support production of microalgae are comprehensively presented and duly discussed, using transport phenomenon and process engineering methodological approaches. The text is subdivided into subsections on: reactor design, which includes tubular reactors, flat plate reactors and fermenter-type reactors; and processing parameters, which include gaseous transfer, medium mixing and light requirements. PMID:17137294

  11. Physical and technical aspects of lead cooled fast reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2001-07-01

    The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

  12. Flexible Architecture System & Topology License Plate Recognition (FAST LPR) and Concept of Operations in Thailand

    OpenAIRE

    Kazantzoglou, Avraam.

    2008-01-01

    This thesis examines the potential that exists in technologically advanced systems to assist the local law enforcement authorities in Thailand in their attempts to continue to effectively control the drug trafficking, despite the advent of newly appeared "threats," like the spread of the amphetamine-type stimulants (ATS). It also provides the pre-Concept of Operations (CONOPS) of such a system, funded by the DoD/CNTPO (Department of Defense/Counter-Narcoterrorism Technology Program Office...

  13. An investigation of integral facility scaling and data relation methods (Integral System Test Program)

    International Nuclear Information System (INIS)

    The Integral Systems Test (IST) Program was initiated in 1982 by government and industry to provide information needed to help resolve issues raised by the accident at the Three Mile Island nuclear power station. Three different integral test facilities, each scaled to a Babcock and Wilcox (B and W) design nuclear steam supply system, will ultimately contribute data to meet the objectives of the program. Each of the facilities was designed using different scaling methodologies, and each has different operating capabilities, such as maximum operating pressure and core power. The overall scaling of each facility is examined in this report, and local scaling is analyzed to demonstrate potential similarities and dissimilarities in facility response relative to expected plant responses. The scaling relationships are used to show how local thermal-hydraulic phenomena in each facility can be compared to each other or to expected plant behavior. The concept of an equilibrium plot is used to show how the global response of each facility can be related for a specific small break loss-of-coolant transient. Potential complications that may arise as a consequence of the facility scaling or facility limitations are enumerated. The potential use of dimensionless groupings for relating and specifying experiments is discussed. Finally, some specific experiments and conditions are proposed for the purpose of simplifying interfacility comparison of test results

  14. Predictions and measurements of isothermal airflow in a model once-through steam generator

    International Nuclear Information System (INIS)

    Once-Through Steam Generators (OTSGs) are used in the Nuclear Steam Supply Systems marketed by The Babcock and Wilcox Company (B and W). To analytically predict the three-dimensional, steady-state thermohydraulic conditions in the OTSG, B and W has developed a proprietary code THEDA-1 and is working in cooperation with EPRI to develop an improved version, THEDA-2. Confident application of THEDA requires experimental verification to demonstrate that the code can accurately describe the thermohydraulic conditions in geometries characteristic of the OTSG. The first step in the THEDA verification process is the subject of this report. A full-scale, partial-section model of two OTSG spans was constructed and tested using isothermal air as the working fluid. Model local velocities and pressure profiles were measured and compared to THEDA prediction for five model configurations. Over 3000 velocity measurements were taken and the results were compared to THEDA predictions. Agreement between measured and predicted velocity data was generally better than +-12.5%

  15. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

  16. The Relation between Experiment and Theory in Reactor Physics

    International Nuclear Information System (INIS)

    The end-product of theoretical and experimental reactor physics work is a detailed knowledge of the neutron physics behaviour of a power reactor, including the reactivity, power distribution and control characteristics. The problem may be divided into two parts. First, the properties of simple geometric configurations of homogeneous or uniform lattice structure and secondly the effects of departures from the idealized system, including the effects of control rods and other absorbers. Alternative approaches to the first problem are possible. For instance, an approximate physical model can be used and experiments carried out covering a range of all the variables. The experiments are then correlated using the simple model and adjusting the nuclear data to give agreement. This approach has been used successfully for natural uranium- fuelled graphite- moderated reactors but is difficult to apply to enriched thermal reactors particularly if more than one fissile material is to be considered. Similarly, for fast reactors good agreement, at least for the smaller systems with relatively hard spectra, can be obtained by ''one group'' calculations in which all neutrons are assumed to have an energy which is associated with a single effective cross-section. The other extreme is to use basic nuclear data and the best physical model which can be conceived with present-day knowledge. Such an approach is made possible by the increased capacity of digital computing machines and by our improved knowledge of nuclear data. The calculations are required to predict the precise physical quantities measured in the experiments. This approach has been used in the liquid moderated reactor field where the experiments are compared with the most elaborate theory and with a theoretical model sufficiently detailed to account for the physical processes involved using basic data, yet adequately rapid in application to be a useful method for design survey purposes. Similar approaches are possible

  17. Thermohydraulic and nuclear modeling of natural fission reactors

    Science.gov (United States)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  18. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  19. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  20. Requirements and expectations from innovative nuclear reactors: Turkey's perspective

    International Nuclear Information System (INIS)

    After the postponement of Akkuyu NPP project in 2000 due to economical reasons, Turkish Atomic Energy Authority (TAEK) commenced a review of national nuclear policy of the country. In the announcement of the postponement Government stated that Turkey's interest to nuclear reactors would continue and Turkey might utilize new generation nuclear reactors in the future. It was also stated that Turkey is willing to participate and contribute to the development of new reactors. In view of these statements and recent developments in the energy sector, TAEK outlined the requirements and expectations regarding new nuclear reactors and decided to participate in some ongoing international studies for the development of innovative reactors. Some of the requirements determined by TAEK are; a) low capital and low electricity generation costs; b) short construction period; c) short licensing period; d) enhanced safety; e) utilization of proven technology; f) environmentally friendly design; g) suitability for public acceptance; h) utilization of indigenous resources; i) and suitability for hydrogen production, desalination and process heat. Studies for the determination of new nuclear policy are continuing. (author)