WorldWideScience

Sample records for northern states monticello reactor

  1. Major plant retrofits at Monticello nuclear generating plant

    International Nuclear Information System (INIS)

    Larsen, D.E.; Hogg, C.B.

    1986-01-01

    For the past several years, Northern States Power (NSP) has been making major plant retrofits to Monticello Nuclear generating Station in order to improve plant availability and upgrade the plant components for the potential extension of the operating license (life extension). This paper discusses in detail three major retrofits that have been completed or in the process of completion; recirculation loop piping replacement, reactor pressure vessel (RPV) water level-instrumentation modification, core spray piping replacement, the authors will address the scope of work, design and installation concerns, and life extension considerations during the design and procurement process for these three projects

  2. The Monticello license renewal project

    International Nuclear Information System (INIS)

    Clauss, J.M.; Harrison, D.L.; Pickens, T.A.

    1993-01-01

    Today, 111 nuclear power plants provide over 20 percent of the electrical energy generated in the United States. The operating license of the oldest operating plant will expire in 2003, one-third of the existing operating licenses will expire by 2010 and the newest plant's operating license will expire in 2033. The National Energy Strategy (NES) prepared by the Department of Energy (DOE) assumes that 70 percent of the current operating plants will continue to operate beyond their current license expiration. Power from current operating plants can assist in ensuring an adequate, diverse, and environmentally acceptable energy supply for economic growth and improved U.S. competitiveness. In order to preserve this energy resource, three major tasks must be successfully completed: (1) establishment of regulations, technical standards, and procedures for the preparation and review of License Renewal Applications (LRAs); (2) development of technical criteria and bases for monitoring, refurbishing or replacing plant equipment; and (3) demonstration of the regulatory process by a plant obtaining a renewed license. Since 1986, the DOE has been working with the nuclear industry and the Nuclear Regulatory Commission (NRC) to establish and demonstrate the option to extend the life of a nuclear power plant by renewing the operating license. The Monticello Lead Plant demonstration project was initiated in September 1988, following the Pilot Plant studies. This paper is primarily focused on the status and insights gained from the Northern States Power Company (NSP) Monticello Lead Plant demonstration project. The following information is included: (1) Current Status - Monticello License Renewal Application; (2) Economic Analysis; (3) License Renewal Regulatory Uncertainty Issues; (4) Key Decisions; (5) Management Structure; (6) Technical and Licensing Perspective; (7) NRC Interactions; (8) Summary

  3. 75 FR 2565 - Northern States Power Company, LLC; Monticello Nuclear Generating Plant Final Environmental...

    Science.gov (United States)

    2010-01-15

    ... impacts. NSPM currently pays annual real estate taxes to Public School District 882, Wright County, and... by plant personnel due to implementation of the proposed EPU. Postulated Accident Doses... significantly. Some of the radioactive waste streams and storage systems evaluated for postulated accidents may...

  4. 75 FR 11578 - Northern States Power Company of Minnesota, Monticello Nuclear Generating Plant; Exemption

    Science.gov (United States)

    2010-03-11

    ...-approved Physical Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Cyber... changes to its security plans. Pursuant to 10 CFR 51.32, ``Finding of no significant impact,'' the... and implementing comprehensive site security programs. The amendments to 10 CFR 73.55 published on...

  5. 75 FR 6224 - Northern States Power Company of Minnesota; Monticello Nuclear Generating Plant Environmental...

    Science.gov (United States)

    2010-02-08

    ... affect radiation exposures to plant workers and members of the public. Therefore, no changes or different... socioeconomic resources. Therefore, no changes to or different types of non-radiological environmental impacts...

  6. 8 x 8 fuel surveillance program at Monticello site - end of Cycle 6: fourth post-irradiation inspection, October 1978

    International Nuclear Information System (INIS)

    Skarshaug, N.H.

    1980-09-01

    A fuel surveillance program for a lead 8 x 8 reload fuel assembly was implemented at the Monticello Nuclear Power Station in May 1974 prior to Reactor Cycle 3. Inspection results of the fourth post-irradiation inspection performed on this surveillance fuel assembly in October 1978 at EOC 6, after a bundle average exposure of 25,900 MWd/MT, are presented. The measurement techniques, results obtained and comparisons to previous measurements are discussed. The bundle and individual rods examined exhibited characteristics of normal operation and were approved for continued irradiation during Monticello operating Cycle 7

  7. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  8. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    International Nuclear Information System (INIS)

    Schlereth, J.R.; Pennington, D.

    1996-01-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it's Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components

  9. Power reactors in member states

    International Nuclear Information System (INIS)

    1975-01-01

    This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

  10. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  11. Oregon State University TRIGA Reactor annual report

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-31

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included.

  12. Oregon State University TRIGA Reactor annual report

    International Nuclear Information System (INIS)

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

    1979-01-01

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included

  13. Biomass utilization at Northern States Power Company

    International Nuclear Information System (INIS)

    Ellis, R.P.

    1994-01-01

    Northern States Power Company (open-quotes NSPclose quotes) generates, transmits and distributes electricity and distributes natural gas to customers in Minnesota, Wisconsin, North Dakota, South Dakota and Michigan. An important and growing component of the fuel needed to generate steam for electrical production is biomass. This paper describes NSP's historical use of biomass, current biomass resources and an overview of how NSP plans to expand its use of biomass in the future

  14. Health assessment for Monticello Radioactive Contaminated Properties, Monticello, Utah, Region 8. CERCLIS No. UTD980667208. Preliminary report

    International Nuclear Information System (INIS)

    1989-01-01

    The Monticello Mill Site (MMS) is on the National Priorities List. The site is located in Monticello (San Juan County), Utah. MMS includes properties contaminated with radioactive ore wastes from a former uranium mill located near the town. The mill was formerly operated by the Atomic Energy Commission and is now the responsibility of the Department of Energy. Radioactive contaminants (uranium, radium, thorium, and radon) are present in mine tailings, soil, ground water, and surface water on-site and also in ground water, surface water, and soil off-site. Based on available information, the site is considered to be of public health concern because of the risk to human health caused by the likelihood of human exposure to hazardous substances. The possibility exists that human exposure could occur from domestic use of contaminated ground water, consumption of commercial crops, garden vegetables grown in contaminated soil, and consumption of commercial livestock that graze on contaminated soil, grasses, and feed

  15. Determination of the probability for radioactive materials on properties in Monticello, Utah

    International Nuclear Information System (INIS)

    Wilson, M.J.; Crutcher, J.W.; Halford, D.K.

    1991-01-01

    The former uranium mill site at Monticello, Utah, is a surplus facility subject to clean-up under the Surplus Facilities Management Program (SFMP). Surrounding properties contaminated with mill site material are also subject to cleanup, and are referred to as Monticello Vicinity Properties (MVP). The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory (ORNL), Grand Junction, Colorado (GJ), was directed by the US Department of Energy (DOE) in July 1988 to assess the radiological condition of properties in Monticello, Utah. Since the Monticello activities are on the National Priority List, extra measures to identify potentially contaminated properties were undertaken. Thus, the likelihood that a random property could contain radioactive materials became a concern to the DOE. The objective of this study was to determine the probability that a vicinity property not addressed under the MVP project could contain Monticello mill-related residual radioactive material in excess of the DOE guidelines. Results suggest approximately 20% of the properties in the Monticello area contain Monticello mill-related residual radioactive material in excess of the DOE guidelines. This suggested that further designation measures be taken prior to the close of the designation phase. A public relations effort that included a property-owner mailing effort, public posting, and newspaper advertisement was one measure taken to ensure that most properties were assessed. As a consequence of this study, DOE directed that radiological screening surveys be conducted on the entirety of the Monticello area

  16. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  17. Executive Summary: Forests of the Northern United States

    Science.gov (United States)

    Stephen R. Shifley; Francisco X. Aguilar; Nianfu Song; Susan I. Stewart; David J. Nowak; Dale D. Gormanson; W. Keith Moser; Sherri Wormstead; Eric J. Greenfield

    2012-01-01

    This executive summary provides an overview of the 200-page report, Forests of the Northern United States, which covers in detail current forest conditions, recent trends, issues, threats and opportunities in the forests in the 20 Northern States. It provides a context for subsequent Northern Forest Futures Project analyses that will forecast alternative future...

  18. Hybrid Reactor designs in the United States

    International Nuclear Information System (INIS)

    Wolkenhauer, W.C.

    1978-01-01

    This paper reviews the current, active, interrelated Hybrid Reactor development programs in the United States, and offers a probable future course of action for the technology. The Department of Energy (DOE) program primarily emphasizes development of Hybrid Reactors that are optimized for proliferation resistance. The Electric Power Research Institute (EPRI) program concentrates on avenues for Hybrid Reactor commercialization. The history of electrical generation technology has been one of steady movement toward higher power densities and higher quality fuels. An apparent advantage of the Hybrid Reactor option is that it follows this trend

  19. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2008-01-01

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

  20. Fluidized Bed Reactor as Solid State Fermenter

    Directory of Open Access Journals (Sweden)

    Krishnaiah, K.

    2005-01-01

    Full Text Available Various reactors such as tray, packed bed, rotating drum can be used for solid-state fermentation. In this paper the possibility of fluidized bed reactor as solid-state fermenter is considered. The design parameters, which affect the performances are identified and discussed. This information, in general can be used in the design and the development of an efficient fluidized bed solid-state fermenter. However, the objective here is to develop fluidized bed solid-state fermenter for palm kernel cake conversion into enriched animal and poultry feed.

  1. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  2. Advanced Reactor Development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Giessing, D. F.; Griffith, J. D.; McGoff, D. J.; Rosen, Sol [U. S. Department of Energy, Texas (United States)

    1990-04-15

    In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE.

  3. Reactor kinetics - pulse and steady state

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B F; Morris, F M [Sandia Laboratories (United States)

    1974-07-01

    An analytical model has been developed which couples the nuclear and thermal characteristics of the Annular Core Pulse Reactor (ACPR) into a solution which describes both the neutron kinetics of the reactor and the temperature behavior of a fuel-moderator element. The model describes both pulse and steady state operations. This paper describes the important aspects of the reactor, the fuel- moderator elements, the neutron kinetic equations of the reactor, and the time-temperature behavior of a fuel-moderator element that is being subjected to the maximum power density in the core. The parameters which are utilized in the equations are divided into two classes, those that can be measured directly and those that are assumed to be known (each is described briefly). Some of the solutions which demonstrate the versatility of the analytical model are described. (author)

  4. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

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

  5. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

    1987-01-01

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  6. Steady-state spheromak reactor studies

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Hagenson, R.L.

    1985-01-01

    After summarizing the essential elements of a gun-sustained spheromak, the potential for a steady-state is explored by means of a comprehensive physics/engineering/costing model. A range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported

  7. Status of the Monticello nuclear generating plant lead plant license renewal program

    International Nuclear Information System (INIS)

    Pickens, T.A.

    1992-01-01

    In 1988, the Monticello nuclear generating plant was chosen by the US Department of Energy through Sandia National Laboratories and the Electric Power Research Institute to serve as the lead boiling water reactor in the lead plant license renewal program. The purpose of the lead plant license renewal program is to provide insights during the development of and to demonstrate the license renewal regulatory process with the US Nuclear Regulatory Commission (NRC). The work being performed in three phases: (1) preparation of the technical basis for license renewal; (2) development of the technical basis into a formal license renewal application; and (3) review of the application by the NRC. This paper discusses the systems and structures identified as important to license renewal in accordance with 10CFR54 as well as the plant documents and programs that were used in going through the identification process. The systems and structures important to license renewal will then provide insights into how structures and components were identified that are required to be evaluated for aging, the elements of the aging evaluations, and the effective programs used to manage potentially significant aging

  8. The Monticello, Utah, uranium mill tailings site: A case history

    International Nuclear Information System (INIS)

    Korte, N.E.; Kearl, P.M.; Sewell, J.M.; Fleischhauer, H.L.; Abramiuk, I.N.

    1984-01-01

    A multidisciplinary study was conducted to characterize the potential for contamination from the inactive millsite in Monticello, Utah. Emphasis was given to site geology, hydrology, and geochemistry for two reasons: (1) a perennial stream flows through the tailings area, and (2) a culinary aquifer is overlain by an alluvial aquifer contaminated by the tailings area. Study results indicate that surface-water contamination attributable to the piles exists for approximately 6 km downstream from the site. Contamination also exists in the alluvial aquifer underlying the millsite. Hydrologic studies indicate an active alluvial system, with recharge to the gravels by infiltration through the trailings. Fortunately, water-level and water-quality data, together with the results of a 51-hour pump test, indicate that the Dakota Formation is an effective aquitard, restricting the downward movement of contaminated water to the underlying culinary aquifer

  9. Investigation of Breast Cancer Risk Factors in northern states of ...

    African Journals Online (AJOL)

    Background: Breast cancer is the most common type of cancers and leading cause of death among women worldwide. In Sudan breast cancer is the most common type of cancer and its incidence has been rising for the past two decades. Objective: To investigate whether the breast risk factors of northern states (Northern ...

  10. Determination of the probability for radioactive materials on properties in Monticello, Utah

    International Nuclear Information System (INIS)

    Wilson, M.J.; Crutcher, J.W.

    1991-02-01

    In 1978, under the authority of the Atomic Energy Act, the US Department of Energy (DOE) established the Surplus Facilities Management Program (SFMP) to manage the maintenance and surveillance of numerous DOE-owned, radioactively contaminated facilities that have been declared surplus and to conduct a program leading to the ultimate disposition of those facilities. The primary responsibility of SFMP is to protect public health and the environment from potentially harmful radioactive contamination contained within or derived from DOE-owned facilities. Management of SFMP is directed by the DOE Office of Environmental Restoration and Waste Management, Washington, DC. Prior to mill site remediation, Monticello properties surrounding the site and designated privately owned are being assessed for inclusion in the SFMP. Oak Ridge National Laboratory (ORNL) was directed by DOE in July 1988 to assess the radiological condition of privately owned properties in Monticello that have been identified as possibly containing Monticello mill-related materials. Properties containing Monticello mill-related materials and with associated radiation levels that exceed US Environmental Protection Agency (EPA) and DOE standards are eligible for cleanup under SFMP. The objective of this study was to determine the probability that a property which contained Monticello mill-related residual radioactive material in excess of the guidelines would not be assessed under the current protocol. 3 refs., 3 figs., 2 tabs

  11. The state of the PIK reactor construction

    International Nuclear Information System (INIS)

    Konoplev, K.A.

    1995-01-01

    Principle concepts of the PIK reactor project were stated late in the 60's but its construction was started in 1976. By the year 1986 the initial project was realised by approximately 70% but then, after Chernobyl accident the construction was essentially frozen to adjust the project to the revised nuclear safety regulations. The revised project was approved only in 1990 when the country was on the threshold of serious economic problems. The PIK reactor is a source of neutrons placed in the heavy water reflector. The fuel is uranium-235 (90% enrichment) of total weight 27 kg. Light water is used as moderator and coolant. Design parameters: thermal power is 100 W; thermal neutron flux in the reflector is 1.2x10 15 n/cm 2 s; in the central vertical beam tube is 5x10 15 n/cm 2 s; number of horizontal beam tubes is 10; diameter of beam tubes is 10 cm, with the possibility of replacement with beam tubes up to 25 cm in diameter. The reactor will be equipped with sources of hot, cold, and ultracold neutrons to obtain beams in different intervals of energy spectrum. The low temperature circuit will enable to irradiate samples at helium temperatures. The reactor has three series cooling circuits. Emergency core cooling systems in LOCA are double and in emergency power supply system is triple. The PIK reactor has no single common containment but four separate systems: for pipelines and units of the first circuit, for heavy water reflector, for operating hall, and for experimental beam tubes hall

  12. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  13. 1980 Environmental monitoring report: US Department of Energy Facilities, Grand Junction, Colorado, and Monticello, Utah

    International Nuclear Information System (INIS)

    1981-04-01

    The effect the Grand Junction, Colorado and Monticello, Utah facilities have on the environment is reflected by the analyses of air, water, and sediment samples. The off-site water and sediment samples were taken to determine what effect the tailings and contaminated equipment buried on the sites may have on the air, water, and adjacent properties

  14. Haematology outreach clinics in the Free State and Northern Cape

    African Journals Online (AJOL)

    patients' domicile, how they were referred, types of diagnoses and ... The intention was to offer a more accessible and affordable service to more patients in the Free State and Northern Cape, in accordance with provincial policy.' The Departments of ... The cost would have been higher if consultants had had to stay over.

  15. Fast reactors fuel Cycle: State in Europe

    International Nuclear Information System (INIS)

    1991-01-01

    In this SFEN day we treat all aspects (economics-reactor cores, reprocessing, experience return) of the LMFBR fuel cycle in Europe and we discuss about the development of this type of reactor (EFR project) [fr

  16. Power reactors in Member States. 1979 edition

    International Nuclear Information System (INIS)

    1979-01-01

    This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

  17. Reactor console replacement at Washington State University

    International Nuclear Information System (INIS)

    Lovas, Thomas A.

    1978-01-01

    A replacement reactor console was installed in 1977 at the W.S.U. 1 MW TRIGA-fueled reactor as the final step in an instrumentation upgrade program. The program was begun circa 1972 with the design, construction and installation of various systems and equipment. Major instruments were installed in the existing console and tested in the course of reactor operation. The culmination of the program was the installation of a cubicle designed and constructed to house the updated instrumentation. (author)

  18. The prevalence of blinding trachoma in northern states of Sudan.

    Directory of Open Access Journals (Sweden)

    Awad Hassan

    Full Text Available BACKGROUND: Despite historical evidence of blinding trachoma, there have been no widespread contemporary surveys of trachoma prevalence in the northern states of Sudan. We aimed to conduct district-level surveys in this vast region in order to map the extent of the problem and estimate the need for trachoma control interventions to eliminate blinding trachoma. METHODS AND FINDINGS: Separate, population based cross-sectional surveys were conducted in 88 localities (districts in 12 northern states of Sudan between 2006 and 2010. Two-stage cluster random sampling with probability proportional to size was used to select the sample. Trachoma grading was done using the WHO simplified grading system. Key prevalence indicators were trachomatous inflammation-follicular (TF in children aged 1-9 years and trachomatous trichiasis (TT in adults aged 15 years and above. The sample comprised 1,260 clusters from which 25,624 households were surveyed. A total of 106,697 participants (81.6% response rate were examined for trachoma signs. TF prevalence was above 10% in three districts and between 5% and 9% in 11 districts. TT prevalence among adults was above 1% in 20 districts (which included the three districts with TF prevalence >10%. The overall number of people with TT in the population was estimated to be 31,072 (lower and upper bounds = 26,125-36,955. CONCLUSION: Trachoma mapping is complete in the northern states of Sudan except for the Darfur States. The survey findings will facilitate programme planning and inform deployment of resources for elimination of trachoma from the northern states of Sudan by 2015, in accordance with the Sudan Federal Ministry of Health (FMOH objectives.

  19. Results of the independent verification of radiological remedial action at 600 South Clayhill Drive (AKA 600 South Cemetery Road), Monticello, Utah (MS00145)

    International Nuclear Information System (INIS)

    Wilson, M.J.; Crutcher, J.W.

    1991-07-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1986 and 1987, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 600 South Cemetery Road (updated by San Juan County and the state of Utah to 600 South Clayhill Drive), Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  20. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  1. A large economic liquid metal reactor for United States utilities

    International Nuclear Information System (INIS)

    Rodwell, E.

    1985-01-01

    The United States has demonstrated its ability to build and operate small and medium sized liquid metal reactors and continues to operate the Experimental Breeder Reactor II and the Fast Flux Test Facility to demonstrate long life fuel designs. Similar-sized liquid metal reactors in Europe have been followed by a step-up to the 1200 MWe capacity of the Superphenix plant. To permit the United States to make a similar step-up in capacity, a 1320 MWe liquid metal reactor plant has been designed with the main emphasis on minimizing the specific capital cost in order to be competitive with light water reactor plant and fossil plant alternatives. The design is based on a four parallel heat transport loops arrangement and complies with current regulatory requirements. The primary heat transport loops are now being integrated into the reactor vessel to achieve further reduction in the capital cost

  2. Conceptual design of the steady state tokamak reactor (SSTR)

    International Nuclear Information System (INIS)

    Oikawa, A.; Kikuchi, M.; Seki, Y.; Nishio, S.; Ando, T.; Ohara, Y.; Takizuka, Tani, K.; Ozeki, T.; Koizumi, K.; Ikeda, B.; Suzuki, Y.; Ueda, N.; Kageyama, T.; Yamada, M.; Mizoguchi, T.; Iida, F.; Ozawa, Y.; Mori, S.; Yamazaki, S.; Kobayashi, T.; Adachi, H.J.; Shinya, K.; Ozaki, A.; Asahara, M.; Konishi, K.; Yokogawa, N.

    1992-01-01

    This paper reports that on the basis of a high bootstrap current fraction observation with JT-60, the concept of steady state tokamak reactor , the SSTR, was conceived and was evolved with the design activity of the SSTR at JAERI. Also results of ITER/FER design activities has enhanced the SSTR design. Moreover the remarkable progress of R and D for fusion reactor engineering, especially in the development of superconducting coils and negative ion based NBI at JAERI have promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor

  3. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  4. 76 FR 44535 - Revisions to the California State Implementation Plan, Northern Sierra Air Quality Management...

    Science.gov (United States)

    2011-07-26

    ... the California State Implementation Plan, Northern Sierra Air Quality Management District, Sacramento Metropolitan Air Quality Management District, and South Coast Air Quality Management District AGENCY... the Northern Sierra Air Quality Management District (NSAQMD), Sacramento Metropolitan Air Quality...

  5. 76 FR 44493 - Revisions to the California State Implementation Plan, Northern Sierra Air Quality Management...

    Science.gov (United States)

    2011-07-26

    ... California State Implementation Plan, Northern Sierra Air Quality Management District, Sacramento Metropolitan Air Quality Management District, and South Coast Air Quality Management District AGENCY... approve revisions to the Northern Sierra Air Quality Management District (NSAQMD), Sacramento Metropolitan...

  6. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  7. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

    Smith, R.R.; Cissel, D.W.

    1978-01-01

    Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO 2 -PuO 2 ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed

  8. Assessment of terrestrial gamma-radiation in Northern State

    International Nuclear Information System (INIS)

    Osman, E. H.

    2007-07-01

    This study is primarily conducted at the request of Northern State government to investigate the present of abnormal radioactivity in Northern State as claimed publicly. Activity concentration of 238 U, 232 Th, 40 K and 137 Cs in soil samples collected from different locations have been measured using high resolution γ-spectrometry. The average concentration were 19±4 Bg/kg for 238 U, 47±11 Bq/kg for 232 Th, 317±65 Bq/kg for 40 K and 2.26 Bq/kg for 137 Cs which means a very little contribution to the total exposure. The obtained results were found to be lower than the corresponding global values reported in the UNSCEAR publication for normal background areas. Absorbed dose rate in air at a height of 1m from the ground was calculated using six sets of dose rate conversion factors and corresponding annual effective dose was estimated. On the average, the values obtained were 52.90, 50.43, 50.41, 43.54, 44.07 and 45.85 nGy.h -1 with corresponding annual effective doses of 64.93, 61.89, 61.87, 53.43, 54.08 and 56.27 μSv/y respectively. These values lie with in the worldwide range for normal radiation areas. The results obtained using these different DRCFs, although the approach used for their estimation was different, revealed no remarkable variation. Using Geographical Information System (GIS), prediction maps for concentration of 238 U, 232 Th, 40 K and 137 Cs was produced. Also a map for absorbed dose rate in air at a height of one-meter above ground level was produced, which showed a trend of increase form the west towards south-east of the State.(Author)

  9. State and parameter estimation in biotechnical batch reactors

    NARCIS (Netherlands)

    Keesman, K.J.

    2000-01-01

    In this paper the problem of state and parameter estimation in biotechnical batch reactors is considered. Models describing the biotechnical process behaviour are usually nonlinear with time-varying parameters. Hence, the resulting large dimensions of the augmented state vector, roughly > 7, in

  10. Dos modos de situarse en el lugar : Monticello de Thomas Jefferson y Taliesin de F. Lloyd Wright

    Directory of Open Access Journals (Sweden)

    Juan Antonio Cortés

    2012-12-01

    Full Text Available

    Resumen

    El artículo consiste en una descripción comparada de dos edificios: Monticello ‐la casa que Thomas Jefferson construyó para sí mismo en Virginia‐ y Taliesin ‐la casa y estudio de Frank Lloyd Wright en Wisconsin‐. El texto estudia en primer lugar las fuentes arquitectónicas y la evolución del proyecto de Monticello, para centrarse después en la explicación de Taliesin. Hay una cierta similitud en el modo en que Monticello y las casas de  Wright ‐en concreto la Ward Willitts‐ se extienden  horizontalmente en el terreno y, volviendo a Taliesin, la tesis  principal del texto es que tanto la residencia de Jefferson como la de Wright se asientan sobre una colina, pero Monticello  ‘corona’ su cima, mientras que Taliesin la bordea, se sitúa como  una ‘ceja’ respecto a la misma. En definitiva, de este  último edificio se puede afirmar que es una ‘casa natural’, que  logra una plena integración entre arquitectura y naturaleza.

    Palabras clave

    casa, proyecto, evolución, corona, natural

    Abstract

    This article consists of a comparative description of two buildings: Monticello, Thomas Jefferson’s residence which he  built for himself in Virginia, and Taliesin, the studio and home of Frank Lloyd Wright in Wisconsin. First, the text studies the  architectural references and evolution of the project for  Monticello, in order to later focus on explaining Taliesin. There is a certain similarity in the way that Monticello and Wright’s  houses (especially the Ward Willits House extend horizontally  across the land. The thesis of this article is that both Jefferson’s and Wright’s residences rest upon a hill, but Monticello crowns  the top while Taliesin borders it like an eyebrow. In conclusion, we can say that Taliesin is a “natural house”, which manages to fully integrate its architecture with nature

  11. The Northern States Power Company welding manual advisor

    International Nuclear Information System (INIS)

    Lu, Yi; Wood, R.M.

    1993-01-01

    The Welding Manual Advisor (WMA) is an object oriented expert system designed to assist Northern States Power (NSP) personnel in implementing the company's Welding Manual. The expert system captures the knowledge of welding experts, addresses important issues in welding activities and automates the use of the Welding Manual. It is estimated that use of the WMA will save $81,000 over the next six years at NSP, because of the reduction of labor and errors in the use of the Welding Manual, and facilitation of training of NSP personnel. The important features of the WMA include the accuracy and consistency in determining welding procedure and requirements, update capability, user friendly interface, on-line help function, back-up capability, and well-documented manuals

  12. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  13. Construction of reactor and outline of breaking state

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji

    1980-01-01

    The Mihama No. 1 reactor of Kansai Electric Power Co., Inc., is the first power-generating PWR in Japan, and it commenced the commercial operation in November, 1970. In June, 1972, the leak of steam generator tubes occurred, and the reactor was stopped for about a half year. In the second periodic inspection in March, 1973, the accident of broken fuel rods was discovered, but it was generally published in December, 1976. In August, 1974, the leak of steam generator tubes occurred again, and since then, the reactor was stopped for a long period. From October, 1978, the reactor has entered the test operation called cycling operation. The Kyoto University Reactor Research Institute lent the cask for transporting the broken pieces of the fuel rods to Kansai Electric Power Co., and obtained the investigation report of Japan Atomic Energy Research Institute, Based on the contract. The accident investigation group of the KURRI has pursued the causes of the accident and examined the propriety of the countermeasures. The outline of the construction of the reactor is described. The upper part of two fuel rods broke and fell on the baffle supporting plate. The broken fuel assembly was named C-34. About a half of fuel pellets and the cladding tube of several tens cm have not yet recovered. The state of break and the presumption of the causes are reported. (Kako, I.)

  14. Support vector machines for nuclear reactor state estimation

    Energy Technology Data Exchange (ETDEWEB)

    Zavaljevski, N.; Gross, K. C.

    2000-02-14

    Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.

  15. Evaluating Russian space nuclear reactor technology for United States applications

    International Nuclear Information System (INIS)

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-01-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch

  16. Support vector machines for nuclear reactor state estimation

    International Nuclear Information System (INIS)

    Zavaljevski, N.; Gross, K. C.

    2000-01-01

    Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm

  17. High Lassa Fever activity in Northern part of Edo State, Nigeria ...

    African Journals Online (AJOL)

    The purpose was to establish simple statistics of the effects of lassa fever in northern part of Edo State, Nigeria. Lassa fever activity in the northern part of Edo state, Nigeria, was confirmed in 2004 by laboratory analysis of samples sent to Bernhard–Nocht Institute (BNI) for Tropical Medicine Hamburg, Germany.

  18. Results of the survey activities and mobile gamma scanning in Monticello, Utah

    International Nuclear Information System (INIS)

    Little, C.A.; Berven, B.A.

    1985-11-01

    The town of Monticello, Utah, was once the site of an active mill which processed vanadium ore (1942 to 1948), and uranium ore (1948 to 1960). Properties in the vicinity of that mill have become contaminated with radioactive material from ore processing. The Radiological Survey Activities (RASA) group at Oak Ridge National Laboratory (ORNL) was requested by the Division of Remedial Action Projects (DRAP) in the Department of Energy (DOE) to: (1) identify potentially contaminated properties; (2) assess natural background radiation levels; and (3) rapidly assess the magnitude, extent, and type (i.e. ore, tailings, etc.) of contamination present on these properties (if any). This survey was conducted by RASA during April 1983. In addition to the 114 properties previously identified from historical information, the ORNL mobile gamma scanning van located 36 new properties exhibiting anomalous gamma radiation levels. Onsite surveys were conducted on 145 of the 150 total properties identified either historically or with the gamma scanning van. Of these 145 properties, 122 of them appeared to have some type of contaminated material present on them; however, only 48 appeared to be contaminated to the extent where they were in excess of Environmental Protection Agency (EPA) criteria (40 CFR 192). Twenty-one other properties were recommended for additional investigation (indoor gamma scanning and radon daughter measurements); of these, only ten required further analysis. This report provides the detailed data and analyses related to the radiological survey efforts performed by ORNL in Monticello, Utah

  19. State of the art of the advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Chawla, R.

    1987-01-01

    A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

  20. Physicochemical state of the spent fuel leaving the reactors

    International Nuclear Information System (INIS)

    Dehaut, Ph.

    2000-01-01

    This report focuses on the current knowledge, updated at the end of 1999, about the physicochemical state of the fuels leaving light water reactors, and particularly pressurized water reactors. Lessons are withdrawn from it making it possible to determine the points which require a necessary deepening of the data and coherence of interpretations. Lastly, evolution of the sailed fuel rod as well as the potential availability of gases and volatile fission products, during a secular storage or of a multi-millennium disposal, are the subject of an attempt at forecast. Accessible data in the scientific literature, or those acquired at the CEA, are particularly numerous. Their analysis and their synthesis are joined together to constitute a collection of references intended to the specialists in nuclear fuel and for all those which contribute to the reflexion on the storage or final disposal of the irradiated fuel. This memory is structured in ten chapters. The last chapter makes it possible to retain on some pages, the essential lessons of this study. Chapter I: Introduction; Chapter II: Characteristics of assemblies and fuels before irradiation; Chapter III: Transformations in reactor; Chapter IV: State of rods leaving the reactor; Chapter V: State of pellets; Chapter VI: Chemical and structural composition of the fuel; Chapter VII: Fuel fragmentation and density; Chapter VIII: Phenomena at the pellet periphery. Formation, characteristics and structure of the rim.Chemical interaction between pellet and cladding; Chapter IX: Location of fission gases and volatile fission products; Chapter X: Review, lessons and predictions. (authors)

  1. Refining aging criteria for northern sea otters in Washington State

    Science.gov (United States)

    Schuler, Krysten L.; Baker, Bridget B.; Mayer, Karl A.; Perez-Heydrich, Carolina; Holahan, Paula M.; Thomas, Nancy J.; White, C. LeAnn

    2018-01-01

    Measurement of skull ossification patterns is a standard method for aging various mammalian species and has been used to age Russian, Californian, and Alaskan sea otter populations. Cementum annuli counts have also been verified as an accurate aging method for the Alaskan sea otter population. In this study, cementum annuli count results and skull ossification patterns were compared as methods for aging the northern sea otter (Enhydra lutris kenyoni) population in Washington State. Significant agreement was found between the two methods suggesting that either method could be used to age the Washington population of otters. This study also found that ossification of the squamosal-jugal suture at the ventral glenoid fossa can be used to differentiate male subadults from adults. To assist field biologists or others without access to cementum annuli or skull ossification analysis techniques, a suite of morphologic, physiologic, and developmental characteristics were analyzed to assess whether a set of these more easily accessible parameters could also predict age class for the Washington population of otters. Tooth condition score, evidence of reproductive activity in females, and tooth eruption pattern were identified as the most useful criteria for classifying Washington sea otters as pups, juveniles, subadults, or adults/aged adults. A simple decision tree based on characteristics accessible in the field or at necropsy was created that can be used to reliably predict age class of Washington sea otters as determined by cementum annuli.

  2. Very high flux steady state reactor and accelerator based sources

    International Nuclear Information System (INIS)

    Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

    2004-01-01

    With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

  3. Steady-state operation requirements of tokamak fusion reactor concepts

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1991-06-01

    In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

  4. State-space representation of the reactor dynamics equations

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1995-01-01

    This paper describes a novel formulation of the reactor space-independent kinetics equations. The intent is to present these equations in a form that is both compatible with modern control theory and mathematically rigorous. It is desired to write the kinetics equations in the standard state variable representation, x = Ax, where x is the state vector and A is the system matrix and, at the same time, avoid mathematical compromises such as the linearization of an equation about a particular operating point. The advantage to this proposed formulation is that it may allow the lateral transfer of existing control concepts, some that have been developed for other fields, to the operation of nuclear reactors. For example, sliding mode control has been developed to allow robots to function in a robust manner in the presence of changes in the system model. This is necessary because a robot is expected to be capable of picking up an object of unknown mass and moving that object along a specified trajectory. The variability of the object's mass introduces an uncertainty into the system model that is used to deduce the appropriate control action. Thus, the robot controller must be made robust against such variations. Sliding mode control is one means of accomplishing this. A reactor controller might benefit from the same concept if its objective were to cause the reactor power to move along a demanded trajectory despite the presence of some uncertainty in the net amount of reactivity that is present

  5. Digital control for the Penn State Breazeale reactor

    International Nuclear Information System (INIS)

    Raiskums, G.A.

    1991-01-01

    Digital control has been an integral part of Canada deuterium uranium (CANDU) nuclear power reactor technology since the 1960s. Much of the high CANDU production reliability can be attributed to the fault-tolerant and flexible control algorithms achievable with digital control. Atomic Energy of Canada Limited (AECL) has now transported this technology to research reactors, using industrial-grade microcomputers to solve equipment aging and spares obsolescence problems so prevalent at older installations. The open architecture of the Intel 8086-based computers provides for wide availability and reasonably priced, quality hardware from numerous sources. AECL recently supplied the Pennsylvania State University Breazeale Reactor (PSBR) with a new console containing a digital control and monitoring system. The reactor safety system (RSS) was also replaced with hardwired relay logic and truly analog state-of-the-art wide range nuclear instrumentation supplied by AECL's subcontractor, Gamma-Metrics. Retaining analog hardware for the mandated RSS functions was key to minimizing licensing efforts and the extensive verification and validation that would be required for safety system software. This paper elaborates on the digital control and monitoring portion of the PSBR console replacement, with emphasis on the key system objectives

  6. MHD stability regimes for steady state and pulsed reactors

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

    1994-02-01

    A tokamak reactor will operate at the maximum value of β≡2μ 0 /B 2 that is compatible with MHD stability. This value depends upon the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near one, I bs /I p ∼ 1, which constrains the product of the inverse aspect ratio and the plasma poloidal beta to be near unity, ε β p ∼ 1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during the ARIES I, II/IV, and III and the PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements on the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies is also discussed

  7. Magnetohydrodynamic stability regimes for steady state and pulsed reactors

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

    1994-01-01

    A tokamak reactor will operate at the maximum value of β≡2μ 0 left angle p right angle /B 2 that is compatible with magnetohydrodynamic (MHD) stability. This value depends on the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near unity, I BS /I P ∼1, which constrains the product of the inverse aspect ratio and the plasma poloidal β to be near unity, arepsilonβ P ∼1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during ARIES I, II/IV, and III and PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements in the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies, is also discussed. ((orig.))

  8. Status of fast breeder reactor development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Rosen, S

    1979-07-01

    This document was prepared by the Office of the Program Director for Nuclear Energy, U.S. Department of Energy (USDOE). It sets forth the status and current activities for the development of fast breeder technology in the United States. In April 1977 the United States announced a change in its nuclear energy policy. Concern about the potential for the proliferation of nuclear weapons capability emerged as a major issue in considering whether to proceed with the development, demonstration and eventual deployment of breeder reactor energy systems. Plutonium recycle and the commercialization of the fast breeder were deferred indefinitely. This led to a reorientation of the nuclear fuel cycle program which was previously directed toward the commercialization of fuel reprocessing and plutonium recycle to the investigation of a full range of alternative fuel cycle technologies. Two major system evaluation programs, the Nonproliferation Alternative Systems Assessment Program (NASAP), which is domestic, and the International Nuclear Fuel Cycle Evaluation (INFCE), which is international, are assessing the nonproliferation advantages and other characteristics of advanced reactor concepts and fuel cycles. These evaluations will allow a decision in 1981 on the future direction of the breeder program. In the interim, the technologies of two fast breeder reactor concepts are being developed: the Liquid Metal Fast Breeder Reactor (LMFBR) and the Gas Cooled Fast Reactor (CFR). The principal goals of the fast breeder program are: LMFBR - through a strong R and D program, consistent with US nonproliferation objectives and anticipated national electric energy requirements, maintain the capability to commit to a breeder option; investigate alternative fuels and fuel cycles that might offer nonproliferation advantages; GCFR - provide a viable alternative to the LMFBR that will be consistent with the developing U.S. nonproliferation policy; provide GCFR technology and other needed

  9. The United States Advanced Reactor Technologies Research and Development Program

    International Nuclear Information System (INIS)

    O’Connor, Thomas J.

    2014-01-01

    The following aspects are addressed: • Nuclear energy mission; • Reactor research development and deployment (RD&D) programs: - Light Water Reactor Sustainability Program; - Small Modular Reactor Licensing Technical Support; - Advanced Reactor Technologies (ART)

  10. COOLOD, Steady-State Thermal Hydraulics of Research Reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-01-01

    1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

  11. Status of fast breeder reactor development in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Horton, K [U.S. Department of Energy, Washington, DC (United States)

    1981-05-01

    The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

  12. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    Horton, K.

    1981-01-01

    The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

  13. Concept study of the Steady State Tokamak Reactor (SSTR)

    International Nuclear Information System (INIS)

    1991-06-01

    The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

  14. Private forest-land owners of the Northern United States, 1994

    Science.gov (United States)

    Thomas W. Birch; Thomas W. Birch

    1996-01-01

    A statistical analytical report on mail canvass of private forest-land owners in the Northern United States. Landowner characteristics attitudes harvesting experience tenure and management planning are discussed.

  15. The United States advanced light water reactor (USALWR) development program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Devine, J.C. Jr.; Sugnet, W.R.

    1987-01-01

    For the United States Nuclear Power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind

  16. The United States Advanced Light Water reactor (USALWR) development program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Devine, Jr.J.C.; Sugnet, W.R.

    1987-01-01

    For the United States Nuclear power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind. (author)

  17. Status of reactor shielding research in the United States

    International Nuclear Information System (INIS)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties

  18. The Current State of Reproductive Health in Rural Northern Nigeria ...

    African Journals Online (AJOL)

    Nigeria has one of the highest maternal mortality ratios in the world, and most deaths occur in the northern part of the country. Concerns about the persistence of the problem prompted some Nigerian academics to partner with their American colleagues to establish a postgraduate fellowship programme that builds the ...

  19. Results of the independent verification of radiological remedial action at 280 South 3rd East Street, Monticello, Utah (MS00099)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Wilson, M.J.

    1990-02-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1985 and 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 280 South 3rd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  20. Results of the independent verification of radiological remedial action at 397 East 3rd South Street, Monticello, Utah (MS00168)

    International Nuclear Information System (INIS)

    Wilson, M.J.; Crutcher, J.W.

    1991-07-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1987 and 1988, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 397 East 3rd South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  1. Results of the independent verification of radiological remedial action at 464 South 1st East Street, Monticello, Utah (MS00071)

    International Nuclear Information System (INIS)

    Wilson, M.J.; Crutcher, J.W.

    1991-07-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1986 and 1987, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 464 South 1st East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  2. Results of the independent verification of radiological remedial action at 273 East 1st South Street, Monticello, Utah (MS00092)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984 and 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 273 East 1st South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  3. Results of the independent verification of radiological remedial action at 98 East 5th South Street, Monticello, Utah (MS00076)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-10-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management program with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 98 East 5th South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  4. Results of the independent verification of radiological remedial action at 381 East 3rd South Street, Monticello, Utah (MS00140)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-10-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity at 381 East 3rd South Street, Monticello, Utah. The Pollutant Assessments group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  5. Results of the independent verification of radiological remedial action at 281 East 3rd South Street, Monticello, Utah (MS00138)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Wilson, M.J.

    1990-02-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 281 East 3rd South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  6. Results of the independent verification of radiological remedial action at 384 South 2nd East Street, Monticello, Utah (MS00084)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-10-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 384 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  7. Results of the independent verification of radiological remedial action at 396 South 2nd East Street, Monticello, Utah (MS00085)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 396 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 34 refs., 2 tabs

  8. Results of the independent verification of radiological remedial action at EG ampersand G Area 6, Monticello, Utah (MS00136)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at EG ampersand G Area 6, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  9. Results of the independent verification of radiological remedial action at 225 South 2nd East Street, Monticello, Utah (MS00114)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1985 and 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 225 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  10. Results of the independent verification of radiological remedial action at 196 South 2nd East Street, Monticello, Utah (MS00135)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 196 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  11. Results of the independent verification of radiological remedial action at 148 East 4th South Street, Monticello, Utah (MS00087)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 148 East 4th South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  12. Results of the independent verification of radiological remedial action at 217 South 2nd East Street, Monticello, Utah (MS00097)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Wilson, M.J.

    1990-02-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1985 and 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 217 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  13. Results of the independent verification of radiological remedial action at 196 East 3rd South Street, Monticello, Utah (MS00083)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-10-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE) Surplus Facilities Management program with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984 and 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 196 East 3rd South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  14. Results of the independent verification of radiological remedial action at 496 South Main Street, Monticello, Utah (MS00050)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-12-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 496 South Main Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  15. Results of the independent verification of radiological remedial action at 87 East 5th South Street, Monticello, Utah (MS00074)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-10-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 87 East 5th South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described

  16. Results of the independent verification of radiological remedial action at 416 South Main Street, Monticello, Utah (MS00150)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Wilson, M.J.

    1990-02-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1984 and 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 416 South Main Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  17. Results of the independent verification of radiological remedial action at 480 South Main Street, Monticello, Utah (MS00049)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Smuin, M.W.

    1989-05-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that has been contaminated by radioactive material resulting from mill operations. During 1984 and 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 480 South Main Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  18. Results of the independent verification of radiological remedial action at 87 East 500 South Street, Monticello, Utah (MS00153)

    International Nuclear Information System (INIS)

    Wilson, M.J.; Crutcher, J.W.

    1991-07-01

    In 1980 the iste of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1987 and 1988, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 87 East 500 South Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  19. Results of the independent verification of radiological remedial action at 433 South 2nd East Street, Monticello, Utah (MS00103)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Wilson, M.J.

    1990-02-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity)properties that had been contaminated by radioactive material resulting from mill operations. During 1984 and 1985, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 433 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 2 tabs., 3 refs

  20. Results of the independent verification of radiological remedial action at 165 North 1st West Street, Monticello, Utah (MS00014)

    International Nuclear Information System (INIS)

    Crutcher, J.W.; Wilson, M.J.

    1990-02-01

    In 1980 the site of a vanadium and uranium mill at Monticello, Utah, was accepted into the US Department of Energy's (DOE's) Surplus Facilities Management Program, with the objectives of restoring the government-owned mill site to safe levels of radioactivity, disposing of or containing the tailings in an environmentally safe manner, and performing remedial actions on off-site (vicinity) properties that had been contaminated by radioactive material resulting from mill operations. During 1985 and 1986, UNC Geotech, the remedial action contractor designated by DOE, performed remedial action on the vicinity property at 217 South 2nd East Street, Monticello, Utah. The Pollutant Assessments Group (PAG) of Oak Ridge National Laboratory was assigned the responsibility of verifying the data supporting the adequacy of remedial action and confirming the site's compliance with DOE guidelines. The PAG found that the site successfully meets the DOE remedial action objectives. Procedures used by PAG are described. 3 refs., 2 tabs

  1. 77 FR 5281 - State-of-the-Art Reactor Consequence Analyses Reports

    Science.gov (United States)

    2012-02-02

    ... NUCLEAR REGULATORY COMMISSION [Docket ID: NRC-2012-0022] State-of-the-Art Reactor Consequence... release of Draft NUREG-1935, ``State-of-the-Art Reactor Consequence Analyses (SOARCA) Report,'' for public... offsite radiological health consequences for potential severe reactor accidents for the Peach Bottom...

  2. Environmental monitoring report on the US Department of Energy's inactive millsite facility, Monticello, Utah, for calendar year 1987

    International Nuclear Information System (INIS)

    1988-05-01

    The inactive Monticello Millsite is located in San Juan County, Utah, just south of the town of Monticello. Environmental monitoring at the site is funded by the Surplus Facilities Management Program (SFMP) and focuses on releases due to preexistent mill tailings. All contaminant discharges result from the leaching of uranium-mill-tailings-related elements by ground water and surface water, and from the release of radon gas and particulate matter into the atmosphere. Pathways facilitating the migration of contaminants from the Monticello site include ground water in the shallow alluvial aquifer underlying the inactive facility, surface water running across the site, and the surrounding atmosphere. Extensive measurement of radon contamination from the tailings piles was conducted during 1984, 1985, and to a lesser extent during 1986 and 1987. On-pile, site-boundary, and off-site atmospheric radon measurements, as well as on- and off-pile radon-flux measurements, were taken. Results of these measurements demonstrate that the EPA standard for radon emissions from inactive uranium processing sites is exceeded at all four tailings piles at the Monticello site. Air particulate monitoring was conducted during 1987 at two on-site locations and at one background location using high-volume Sierra-Anderson model 300 air particulate samplers. So that only the inhalable particles would be collected, 10-micron-size screens were added to the samplers. The maximum airborne concentrations of radium-226, thorium-230, and uranium were all several orders of magnitude below the regulatory limits specified by DOE Order 5480.1. 22 refs., 5 figs., 9 tabs

  3. Nuclear power reactor licensing and regulation in the United States

    International Nuclear Information System (INIS)

    Shapar, H.K.

    1979-01-01

    The report is devoted to four subjects: an explanation of the origins, statutory basis and development of the present regulatory system in the United States; a description of the various actions which must be taken by a license applicant and by the Nuclear Regulatory Commission before a nuclear power plant can be constructed and placed on-line, an account of the current regulatory practices followed by the US NRC in licensing nuclear power reactors; an identification of some of the 'lessons learned' from the Three Mile Island accident and some proposed regulatory and legislative solutions. (NEA) [fr

  4. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.

    2001-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  5. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.; Voitsekhovitch, I.

    1999-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  6. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  7. Futures project anticipates changes and challenges facing forests of the northern United States

    Science.gov (United States)

    Stephen R. Shifley; W. Keith Moser; Michael E. Goerndt; Nianfu Song; Mark D. Nelson; David J. Nowak; Patrick D. Miles; Brett J. Butler; Ryan D. DeSantis; Francisco X. Aguilar; Brian G. Tavernia

    2014-01-01

    The Northern Forest Futures Project aims to reveal how today's trends and choices are likely to change the future forest landscape in the northeastern and midwestern United States. The research is focused on the 20-state quadrant bounded by Maine, Maryland, Missouri, and Minnesota. This area, which encompasses most of the Central Hardwood Forest region, is the...

  8. Monticello Mill Tailings Site, Operable Unit lll, Annual Groundwater Report, May 2015 Through April 2016

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [USDOE Office of Legacy Management (LM), Washington, DC (United States); Smith, Fred [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-10-01

    This report provides the annual analysis of water quality restoration progress, cumulative through April 2016, for Operable Unit (OU) III, surface water and groundwater, of the U.S. Department of Energy (DOE) Office of Legacy Management (LM) Monticello Mill Tailings Site (MMTS). The MMTS is a Comprehensive Environmental Response, Compensation, and Liability Act National Priorities List site located in and near the city of Monticello, San Juan County, Utah. MMTS comprises the 110-acre site of a former uranium- and vanadium-ore-processing mill (mill site) and 1700 acres of surrounding private and municipal property. Milling operations generated 2.5 million cubic yards of waste (tailings) from 1942 to 1960. The tailings were impounded at four locations on the mill site. Inorganic constituents in the tailings drained from the impoundments to contaminate local surface water (Montezuma Creek) and groundwater in the underlying alluvial aquifer. Mill tailings dispersed by wind and water also contaminated properties surrounding and downstream of the mill site. Remedial actions to remove and isolate radiologically contaminated soil, sediment, and debris from the former mill site, Operable Unit I (OU I), and surrounding properties (OU II) were completed in 1999 with the encapsulation of the wastes in an engineered repository located on DOE property 1 mile south of the former mill site. This effectively removed the primary source of groundwater contamination; however, contamination of groundwater and surface water remains within OU III at levels that exceed water quality protection standards. Uranium is the primary contaminant of concern (COC). LM implemented monitored natural attenuation with institutional controls as the OU III remedy in 2004. Because groundwater restoration proceeded more slowly than expected and did not meet performance criteria established in the OU III Record of Decision (June 2004), LM implemented a contingency action in 2009 by an Explanation of

  9. 77 FR 46008 - Approval and Promulgation of State Implementation Plans: Idaho; Boise-Northern Ada County Air...

    Science.gov (United States)

    2012-08-02

    ... Promulgation of State Implementation Plans: Idaho; Boise-Northern Ada County Air Quality Maintenance Area... the Northern Ada County Air Quality Maintenance Area Second 10-year Carbon Monoxide Maintenance Plan...-Northern Ada County Air Quality Maintenance Area will maintain air quality standards for carbon monoxide...

  10. Continuous cryopump for steady state mirror fusion reactors

    International Nuclear Information System (INIS)

    Batzer, T.H.; Call, W.R.

    1983-01-01

    The characteristics of mirror fusion reactors, i.e., steady state operation, a low neutral gas density, and a large gas throughput require unique vacuum pumping capabilities. One approach that appears to meet these requirements is a liquid helium-cooled cryopump system in which a fixed portion can be isolated and degassed while the remainder continues to pump. The time to degas a rotating, fixed portion of the pumping area and the ratio of that area to the total area fixes the gas inventory in the chamber. It follows that the active pump area maintains the required neutral gas density and the time-averaged degassing rate equals the gas throughput. We have built such a cryopump whereby the gas condensed (deuterium) on the liquid helium-cooled panel can be transferred to a collector pump and subsequently to an exterior mechanical pump and exhausted. At panel loadings as high as 0.55 Torr-/lcm 2 the gas leakage during degassing is less than 8% and the degassing time is less than 10 min. Scaling to reactor size appears to be feasible

  11. Fast reactors in Russia: State of the art and trends of development

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Ashurko, Yu.M.; Zverev, K.V.; Oshkanov, N.N.; Korol'kov, A.S.; Filin, A.I.

    2002-01-01

    This status report contains the following: facts on nuclear power in Russia from 2001-2002; plans for further development of nuclear power; state of the art on operation of fast reactors in 2002, namely BN-600, experimental reactors BOR-60 and BR-10; construction of NPP BN-800; participation in activities on BN-350 reactor decommissioning; description of trends of design studies in the field of fast reactors and accelerator driven systems

  12. BR2 reactor core steady state transient modeling

    International Nuclear Information System (INIS)

    Makarenko, A.; Petrova, T.

    2000-01-01

    A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

  13. Solid state laser driver for an ICF reactor

    International Nuclear Information System (INIS)

    Krupke, W.F.

    1988-01-01

    A conceptual design is presented of the main power amplifier of a multi-beamline, multi-megawatt solid state ICF reactor driver. Simultaneous achievement of useful beam quality and high average power is achieved by a proper choice of amplifier geometry. An amplifier beamline consists of a sequence of face-pumped rectangular slab gain elements, oriented at the Brewster angle relative to the beamline axis, and cooled on their large faces by helium gas that is flowing subsonically. The infrared amplifier output radiation is shifted to an appropriately short wavelength ( 10% (including all flow cooling input power) when the amplifiers are pumped by efficient high-power AlGaAs semiconductor laser diode arrays. 11 refs., 3 figs., 7 tabs

  14. Simulation of pressurized water reactor in accidental state

    International Nuclear Information System (INIS)

    Chakir, E.

    1994-01-01

    The aim of this work is to develop the 1300 MWe 4 loops 'PWR' simulator called 'SATRAPE', witch the adopted physics modelisation allows a simplified neutronic calculation, and focus essentially on the reactor thermal hydraulic behavior in the case of the following accidents: - Loss of Coolant Accident (LOCA). - Steam Generator Tube Failure (SGTF). - Steam Line Break (SLB). In case of the 'LOCA' or 'SLB' accident, this modelisation enables the calculation of the pressure and the temperature in the containment building, and also the debit of the released dose in this latter in case of the 'LOCA' accident. The adopted models are relatively simple so as to allow an explicit resolve. In SATRAPE, two graphical interfaces enables to launch orders, whereas the other permits to visualize, the principal state variables of installations. The results obtained show a very good consistency with the envisaged commonly scenario at the time of the considered accidents. 33 refs., 52 figs., 1 tab. (author)

  15. Estimates of vertical hydraulic conductivity in the middle Dakota Sandstone, Monticello, Utah

    International Nuclear Information System (INIS)

    Kautsky, M.; Kearl, P.M.; Dexter, J.J.; Zinkl, R.J.

    1986-01-01

    There are about 2 million tons of uranium mill tailings which lie directly on top of an alluvial aquifer at the Monticello millsite, Utah. The aquifer is contaminated as a consequence of leachate percolating through the tailings. The Burro Canyon Formation which is the local culinary aquifer, underlies the site at depth, but is isolated from the alluvial aquifer by an aquitard composed primarily of middle Dakota Sandstone, and some Mancos Shale. Water quality monitoring of the Burro Canyon aquifer has indicated that it contains very low to no contamination by radionuclides. Tritium data have shown that the recharge to the aquifer predates 1953. Pump tests conducted on the system using the ratio method, have shown the vertical hydraulic conductivity of the aquitard is some 5.2 x 10/sup -7/ to 8.0 x 10/sup -9/ m/d (1.7 x 10/sup -6/ to 2.6 x 10/sup -8/ ft/d). Based upon the aquifer monitoring and test data, the middle Dakota Sandstone appears to be an effective aquitard impeding the downward migration of contaminants from the alluvial aquifer to the Burro Canyon aquifer

  16. Evaluation of Monticello Nuclear Power Plant, Environmental Impact Prediction, based on monitoring programs

    International Nuclear Information System (INIS)

    Gore, K.L.; Thomas, J.M.; Kannberg, L.D.; Watson, D.G.

    1976-11-01

    This report evaluates quantitatively the nonradiological environmental monitoring programs at Monticello Nuclear Generating Plant. The general objective of the study is to assess the effectiveness of monitoring programs in the measurement of environmental impacts. Specific objectives include the following: (1) Assess the validity of environmental impact predictions made in the Environmental Statement by analysis of nonradiological monitoring data; (2) evaluate the general adequacy of environmental monitoring programs for detecting impacts and their responsiveness to Technical Specifications objectives; (3) assess the adequacy of preoperational monitoring programs in providing a sufficient data base for evaluating operational impacts; (4) identify possible impacts that were not predicted in the environmental statement and identify monitoring activities that need to be added, modified or deleted; and (5) assist in identifying environmental impacts, monitoring methods, and measurement problems that need additional research before quantitative predictions can be attempted. Preoperational as well as operational monitoring data were examined to test the usefulness of baseline information in evaluating impacts. This included an examination of the analytical methods used to measure ecological and physical parameters, and an assessment of sampling periodicity and sensitivity where appropriate data were available

  17. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    accidents and, secondly, the physical phenomena, studies and analyses described in Chapters 5 to 8. Chapter 5 is devoted to describing the physical phenomena liable to occur during a core melt accident, in the reactor vessel and the reactor containment. It also presents the sequence of events and the methods for mitigating their impact. For each of the subjects covered, a summary of the physical phenomena involved is followed by a description of the past, present and planned experiments designed to study these phenomena, along with their modelling, the validation of which is based on the test results. The chapter then describes the computer codes that couple all of the models and provide the best current state of knowledge of the phenomena. Lastly, this knowledge is reviewed while taking into account the gaps and uncertainties, and the outlook for the future is presented, notably regarding experimental programmes and the development of modelling and numerical simulation tools. Chapter 6 focuses on the behaviour of the containment enclosures during a core melt accident. After summarising the potential leakage paths of radioactive substances through the different containments in the case of the accidents chosen in the design phase, it presents the studies of the mechanical behaviour of the different containments under the loadings that can result from the hazards linked with the phenomena described in Chapter 5. Chapter 6 also discusses the risks of containment building bypass in a core melt accident situation. Chapter 7 presents the lessons learned regarding the phenomenology of core melt accidents and the improvement of nuclear reactor safety. Lastly, Chapter 8 presents a review of development and validation efforts regarding the main computer codes dealing with 'severe accidents', which draw on and build upon the knowledge mainly acquired through the research programmes: ASTEC (IRSN and GRS), MAAP-4 (FAI (US)) and used by EDF and by utilities in many other

  18. War and early state formation in the northern Titicaca Basin, Peru.

    Science.gov (United States)

    Stanish, Charles; Levine, Abigail

    2011-08-23

    Excavations at the site of Taraco in the northern Titicaca Basin of southern Peru indicate a 2,600-y sequence of human occupation beginning ca. 1100 B.C.E. Previous research has identified several political centers in the region in the latter part of the first millennium B.C.E. The two largest centers were Taraco, located near the northern lake edge, and Pukara, located 50 km to the northwest in the grassland pampas. Our data reveal that a high-status residential section of Taraco was burned in the first century A.D., after which economic activity in the area dramatically declined. Coincident with this massive fire at Taraco, Pukara adopted many of the characteristics of state societies and emerged as an expanding regional polity. We conclude that organized conflict, beginning approximately 500 B.C.E., is a significant factor in the evolution of the archaic state in the northern Titicaca Basin.

  19. On-line computer control of a nuclear reactor using optimal control and state estimation methods

    International Nuclear Information System (INIS)

    Tye, C.

    1980-01-01

    This paper describes the experimental implementation of a nuclear reactor control system using combined optimal state feedback based on the Quadratic Regulator and state estimation using Kalman filtering techniques. The results obtained from the experiments indicate that a reactor control loop designed using this approach has improved stability margins, greater speed of response and noise filtering properties compared with a conventional reactor control loop. 11 refs

  20. Annual report on the state of RB reactor components and equipment, december 1999

    International Nuclear Information System (INIS)

    Milosevic, M.

    1999-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 13 times, meaning that experiments were done with 13 configurations of the reactor core. Total reactor operation amounted to 84 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1999, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  1. Annual report on the state of RB reactor components and equipment, december 1998

    International Nuclear Information System (INIS)

    Milosevic, M.

    1998-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  2. Review of tokamak power reactor and blanket designs in the United States

    International Nuclear Information System (INIS)

    Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

    1986-01-01

    The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

  3. Data Validation Package October 2015 Groundwater and Surface Water Sampling at the Monticello, Utah, Processing Site January 2016

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [U.S. Dept. of Energy, Washington, DC (United States). Office of Legacy Management; Smith, Fred [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2016-01-21

    Sampling Period: October 12–14, 2015. This semiannual event includes sampling groundwater and surface water at the Monticello Mill Tailings Site. Sampling and analyses were conducted as specified in the 2004 Monticello Mill Tailings Site Operable Unit III Post-Record of Decision Monitoring Plan, Draft Final and Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). Samples were collected from 52 of 61 planned locations (15 of 17 former mill site wells, 17 of 18 downgradient wells, 9 of 9 downgradient permeable reactive barrier wells, 2 of 7 seeps and wetlands, and 9 of 10 surface water locations). Locations MW00-07, Seep 1, Seep 2, Seep 3, Seep 5, Seep 6, SW00-01, T01-13, and T01-19 were not sampled because of insufficient water availability. All samples were filtered as specified in the monitoring plan. Duplicate samples were collected from surface water location W3-04 and from monitoring wells 82-08, 92-09, and 92-10. Water levels were measured at all but one sampled well and an additional set of wells. The contaminants of concern (COCs) for the Monticello Mill Tailings Site are arsenic, manganese, molybdenum, nitrate + nitrite as nitrogen (nitrate + nitrite as N), selenium, uranium, and vanadium. Time-concentration graphs of the COCs for all groundwater and surface water locations are included in this report. Locations with COCs that exceeded remediation goals are listed.

  4. Opportunities and challenges related to the development of small modular reactors in mines in the Northern Territories of Canada

    Energy Technology Data Exchange (ETDEWEB)

    Sam-Aggrey, H., E-mail: godfree17@hotmail.com [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Small modular reactors (SMRs) are being touted as safer, more cost effective, and more flexible than traditional nuclear power plants. Consequently, it has been argued that SMR technology is pivotal to the revitalization of the nuclear industry at the national and global levels. Drawing mainly on previously published literature, this paper explores the opportunities and challenges related to the deployment of SMRs in the northern territories of Canada. The paper examines the potential role of SMRs in providing an opportunity for remote mines in northern Canada to reduce their vulnerability and dependence on costly, high-carbon diesel fuel. The paper also outlines and discusses some of the potential socio-economic barriers that could impede the successful introduction of SMRs in the territories. These issues include: economic factors (such as the price of primary minerals and economics of mineral exploration, and the cost of SMR deployment), the lack of infrastructure in the territories to support mining developments, and the issues pertaining to the social acceptance of nuclear power generation. (author)

  5. Maternal Mortality At The State Specialist Hospital Bauchi, Northern ...

    African Journals Online (AJOL)

    Objective: To analyse and document our experiences with maternal mortality with the view of finding the trends over the last seven years, common causes and attributing socio-demographic factors. Design: A prospective analysis of maternal mortality. Setting: State Specialists Hospital Bauchi, Bauchi Northeastern Nigeria.

  6. Effects of loading reactivity at dynamic state on wave of neutrons in burst reactor

    International Nuclear Information System (INIS)

    Gao Hui; Liu Xiaobo; Fan Xiaoqiang

    2013-01-01

    Based on the point reactor model, the program for simulating the burst of reactors, including delay neutron, thermal feedback and reactivity of rod, was developed. The program proves to be suitable to burst reactor by experimental data. The program can describe the process of neutron-intensity change in burst reactors. With the program, the parameters of burst (wave of burst, power of peak and reactivity of reactor) under the condition of dynamic reactivity can be calculated. The calculated result demonstrates that the later the burst is initiated, the greater its power of peak and yield are and that the maximum yield coordinates with the yield under static state. (authors)

  7. Data Validation Package October 2016 Groundwater and Surface Water Sampling at the Monticello, Utah, Disposal and Processing Sites January 2017

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [USDOE Office of Legacy Management (LM), Washington, DC (United States); Smith, Fred [Navarro Research and Engineering, Inc., Grand Junction, CO (United States)

    2017-02-01

    Sampling Period: October 10–12, 2016. This semiannual event includes sampling groundwater and surface water at the Monticello Disposal and Processing Sites. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated) and Program Directive MNT-2016-01. Samples were collected from 54 of 64 planned locations (16 of 17 former mill site wells, 15 of 18 downgradient wells, 7 of 9 downgradient permeable reactive barrier wells, 3 of 3 bedrock wells, 4 of 7 seeps and wetlands, and 9 of 10 surface water locations).

  8. Timber harvesting patterns for major states in the central, northern, and mid-Atlantic hardwood regions

    Science.gov (United States)

    William G. Luppold; Matthew S. Bumgardner

    2018-01-01

    Timber harvesting is a major disturbance agent influencing the composition and structure of eastern hardwood forests. To better understand timber harvesting practices, we examined roundwood harvesting patterns in 13 eastern states in the Central, Mid-Atlantic, and Northern regions that contained high proportional volumes of hardwood in their forest inventories. Nearly...

  9. Reactor aging research. United States Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    Vassilaros, M.G.

    1998-01-01

    The reactor ageing research activities in USA described, are focused on the research of reactor vessel integrity, including regulatory issues and technical aspects. Current emphasis are described for fracture analysis, embrittlement research, inspection capabilities, validation od annealing rule, revision of regulatory guide

  10. The state of art report on advanced reactor development

    International Nuclear Information System (INIS)

    Kim, Keung Koo; Noh, J. M.; Hwang, D. H. and others

    1999-07-01

    Recently, researches on the advanced power reactors are being performed actively, that maximize the economics and enhance the reactor safety by introducing the inherent safety characteristics and passive safety features. In the development of advanced reactor technology, we developed the inherent core design technologies which can form a foundation of indigenous technologies to provide the basic technology for the core design of the domestic advanced reactor. In this report, we examined the neutronics design technologies and core thermal hydraulics design technologies for advanced reactors performed all over the world. Major efforts are focussed on the soluble boron free core design technology and high conversion core design technology. In addition to these, new conceptual core, such as a supercritical core, design technology development was also reviewed. The characteristics of critical heat flux have been investigated for non-square lattice rod bundles, such as triangular lattice and wire wrap lattice. Based on the status of advanced reactor development, the soluble boron free and hexagonal lattice core design technologies are elementary technology for the domestic advanced reactor core. These elementary core technologies would enhance the reactor safety and improve the economics. (author). 71 refs., 31 tabs., 74 figs

  11. Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

    International Nuclear Information System (INIS)

    Azpitarte, O.; Ramilo, L.

    2013-01-01

    The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

  12. Ostrich Management practices in three states of Northern Nigeria

    Directory of Open Access Journals (Sweden)

    Mshelia

    2011-04-01

    Full Text Available The study was conducted to identify management practices associated with ostrich farming in Kano, Kaduna and Plateau States of Nigeria. Seven farms were purposively selected as units of analysis. Primary data were generated by means of a pre-tested, semi-structured questionnaire, administered to the sampled respondents. A simple inductive statistics was applied to the primary data. The result reveals a commercial production of ostrich by 86 % with all the farms engaged in production of other livestock species. Similarly, all the farms had shelter for chicks and breeders which were all erected using wire mesh and poles at above 5 feet fencing level. More over, 100 % of the farms were densely stocked (below 500 m2 for a pair of ostrich with facilities below recommended levels. The result also showed that 100 % of the farms compound feed locally using premix without no providing grit and low (14 % usage of succulent feed. About 29 % of the farms use endoparasitic and anthelminthic drugs as health management practices. On the reproductive practices, only trio (29 % and colony (71 % configurations were practiced. The prominent biosecurity measures include division of farm into disease control unit (100 % and employee enlightenment (86 %. [Veterinary World 2011; 4(2.000: 64-67

  13. Radionuclide concentrations in the northern part of The Netherlands after the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    de Meijer, R.J.; Aldenkamp, F.J.; Brummelhuis, M.J.; Jansen, J.F.; Put, L.W.

    1990-01-01

    Concentrations of radionuclides originating from the Chernobyl reactor accident were measured as a function of time in air, rainwater, grass, cow's milk, vegetables and dust by means of high-resolution gamma-ray spectroscopy. Special attention was paid to grass and milk originating from the same meadows. Also, milk of cows temporarily kept inside after the accident was monitored until a few days after their release from the stables. Activity ratios in various types of samples and the implication of the sheltering measures for cows are discussed

  14. Pennsylvania State University Breazeale Nuclear Reactor. Thirtieth annual progress report, July 1, 1984-June 30, 1985

    International Nuclear Information System (INIS)

    Levine, S.H.; Totenbier, R.E.

    1985-08-01

    This report is the thirtieth annual progress report of the Pennsylvania State University Breazeale Nuclear Reactor and covers such topics as: personnel; reactor facility; cobalt-60 facility; education and training; Radionuclear Application Laboratory; Low Level Radiation Monitoring Laboratory; and facility research utilization

  15. Main results of BN-600 reactor stress-strain state investigations

    International Nuclear Information System (INIS)

    Panov, V.A.

    1983-01-01

    The development of BN-600 fast reactor plant needed the solution of a series of complex engineering problems including ones for confirming integrity of the most vital structural components. The particular attention was given to the main vessel since reactor availability end safe operation of the plant as a whole depend on vessel strength end integrity. The present report deals with the main results of theoretical and experimental investigations of the stress-strain state of BN-600 reactor vessel carried out during design, start-up and initial bringing the reactor to power

  16. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  17. 76 FR 26224 - Revisions to the California State Implementation Plan, Northern Sonoma County Air Pollution...

    Science.gov (United States)

    2011-05-06

    ...EPA is proposing to approve revisions to the Northern Sonoma County Air Pollution Control District (NSCAPCD) and Mendocino County Air Quality Management District (MCAQMD) portions of the California State Implementation Plan (SIP). Both districts are required under Part C of title I of the Clean Air Act (CAA) to adopt and implement SIP- approved Prevention of Significant Deterioration (PSD) permit programs. These proposed revisions update the definitions used in the districts' PSD permit programs.

  18. 76 FR 26192 - Revisions to the California State Implementation Plan, Northern Sonoma County Air Pollution...

    Science.gov (United States)

    2011-05-06

    ...EPA is taking direct final action to approve revisions to the Northern Sonoma County Air Pollution Control District (NSCAPCD) and Mendocino County Air Quality Management District (MCAQMD) portions of the California State Implementation Plan (SIP). Both districts are required under Part C of title I of the Clean Air Act (CAA) to adopt and implement SIP-approved Prevention of Significant Deterioration (PSD) permit programs. These revisions update the definitions used in the districts' PSD permit programs.

  19. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  20. Practitioner survey of the state of health integration in environmental assessment: The case of northern Canada

    International Nuclear Information System (INIS)

    Noble, Bram; Bronson, Jackie

    2006-01-01

    Based on a case study of health integration in Canadian northern EA, this paper further demonstrates the lack of consistent integration of health in EA practice. A survey was administered to northern EA and health practitioners, administrators and special interest groups to assess current northern health assessment practices, the scope of health in EA, EA performance with regard to health assessment and the perceived barriers to health integration. Results suggest that health is currently recognized as an important component of northern EA and is addressed in the majority of cases; however, health is addressed primarily during the pre-decision stages of EA and less often during post-decision follow-up and monitoring. Moreover, when health is addressed, attention is limited to the physical components of health and health impacts due to physical environmental change, with considerably less attention given to the social aspects of health. Results also suggest dissent between EA practitioners, health practitioners and other interests concerning the overall state of health in EA; however, there is consensus on the key challenges to improved integration, namely differences in understanding of the scope of health and expectations of EA to assess health impacts; limited coordination between EA and health practitioners; limited scope and requirements of current EA legislation for health assessment; and the lack of supporting EA methods and frameworks

  1. [Poles of American tegumentary leishmaniasis production in northern Paraná State, Brazil].

    Science.gov (United States)

    Monteiro, Wuelton Marcelo; Neitzke, Herintha Coeto; Silveira, Thaís Gomes Verzignassi; Lonardoni, Maria Valdrinez Campana; Teodoro, Ueslei; Ferreira, Maria Eugênia Moreira Costa

    2009-05-01

    American tegumentary leishmaniasis is endemic in the State of Paraná, with 99.3% of the cases reported in the South of Brazil. Spatial distribution of the disease in northern Paraná was verified, identifying the most relevant geographic areas in epidemiological terms. The study used data recorded on epidemiological forms from the Teaching and Research Clinical Test Laboratory of the State University in Maringá, from 1987 to 2004. The study only included individuals that were infected in the municipalities (counties) in northern Paraná. Identification of the epidemiological units (poles and circuits) was based on spatial density of cases, according to the model proposed by the National Health Foundation, considering the most likely infection sites. Considering 1,933 reported cases, 1,611 were infected in northern Paraná. American tegumentary leishmaniasis distribution in Paraná State suggests two circuits for production of the disease: Paraná-Paranapanema, highlighting the Cinzas-Laranjinha, Tibagi, Ivaí-Pirapó, Piquiri, and Baixo Iguaçu poles, and Ribeira, highlighting the Alto Ribeira pole.

  2. 77 FR 45962 - Approval and Promulgation of State Implementation Plans: Idaho; Boise-Northern Ada County Air...

    Science.gov (United States)

    2012-08-02

    ... Promulgation of State Implementation Plans: Idaho; Boise-Northern Ada County Air Quality Maintenance Area... (IDEQ) submitted the Northern Ada County Air Quality Maintenance Area Second 10-year Carbon Monoxide... Ada County Air Quality Maintenance Area will maintain air quality standards for carbon monoxide (CO...

  3. The reactor trainer: state-of-the-art classroom learning

    International Nuclear Information System (INIS)

    Stater, R.G.

    1996-01-01

    The Reactor Trainer is a professional, PC based, graphically enhanced, training resource specifically developed and customized for Class Room teaching of and learning about, reactor behavior. This unique, and focused, learning-target sets The Trainer apart from the panorama of the more common PC plant simulator. Its educational scope extends along a logical learning path, starting with important fundamental behavioral concepts of delayed neutrons, neutron multiplying factors, and reactor rate, moving to simple reactor transients in real time, and culminating with more complex operational evolutions. The Trainer empowers the Instructor with a dynamic Class Room demonstrator and the student with a superior hands-on learning tool. The Trainer's versatility encompasses a wide variety of educational needs, including initial operator training, requalification training, Shift Technical Advisor training, and other advanced or specialized training. In addition, The Reactor Trainer enhances prerequisite preparation of operator candidates for full-scale control room training and, in so doing, PC economics relieves full-scale simulator hours. (author)

  4. Geochemical orientation survey of stream sediment, stream water, and ground water near uranium prospects, Monticello area, New York. National Uranium Resource Evaluation Program

    International Nuclear Information System (INIS)

    Rose, A.W.; Smith, A.T.; Wesolowski, D.

    1982-08-01

    A detailed geochemical test survey has been conducted in a 570 sq km area around six small copper-uranium prospects in sandstones of the Devonian Catskill Formation near Monticello in southern New York state. This report summarizes and interprets the data for about 500 stream sediment samples, 500 stream water samples, and 500 ground water samples, each analyzed for 40 to 50 elements. The groundwater samples furnish distinctive anomalies for uranium, helium, radon, and copper near the mineralized localities, but the samples must be segregated into aquifers in order to obtain continuous well-defined anomalies. Two zones of uranium-rich water (1 to 16 parts per billion) can be recognized on cross sections; the upper zone extends through the known occurrences. The anomalies in uranium and helium are strongest in the deeper parts of the aquifers and are diluted in samples from shallow wells. In stream water, copper and uranium are slightly anomalous, as in an ore factor derived from factor analysis. Ratios of copper, uranium, and zinc to conductivity improve the resolution of anomalies. In stream sediment, extractable uranium, copper, niobium, vanadium, and an ore factor furnish weak anomalies, and ratios of uranium and copper to zinc improve the definition of anomalies. The uranium/thorium ratio is not helpful. Published analyses of rock samples from the nearby stratigraphic section show distinct anomalies in the zone containing the copper-uranium occurrences. This report is being issued without the normal detailed technical and copy editing, to make the data available to the public before the end of the National Uranium Reconnaissance Evaluation program

  5. Geochemical orientation survey of stream sediment, stream water, and ground water near uranium prospects, Monticello area, New York. National Uranium Resource Evaluation Program

    Energy Technology Data Exchange (ETDEWEB)

    Rose, A. W.; Smith, A. T.; Wesolowski, D.

    1982-08-01

    A detailed geochemical test survey has been conducted in a 570 sq km area around six small copper-uranium prospects in sandstones of the Devonian Catskill Formation near Monticello in southern New York state. This report summarizes and interprets the data for about 500 stream sediment samples, 500 stream water samples, and 500 ground water samples, each analyzed for 40 to 50 elements. The groundwater samples furnish distinctive anomalies for uranium, helium, radon, and copper near the mineralized localities, but the samples must be segregated into aquifers in order to obtain continuous well-defined anomalies. Two zones of uranium-rich water (1 to 16 parts per billion) can be recognized on cross sections; the upper zone extends through the known occurrences. The anomalies in uranium and helium are strongest in the deeper parts of the aquifers and are diluted in samples from shallow wells. In stream water, copper and uranium are slightly anomalous, as in an ore factor derived from factor analysis. Ratios of copper, uranium, and zinc to conductivity improve the resolution of anomalies. In stream sediment, extractable uranium, copper, niobium, vanadium, and an ore factor furnish weak anomalies, and ratios of uranium and copper to zinc improve the definition of anomalies. The uranium/thorium ratio is not helpful. Published analyses of rock samples from the nearby stratigraphic section show distinct anomalies in the zone containing the copper-uranium occurrences. This report is being issued without the normal detailed technical and copy editing, to make the data available to the public before the end of the National Uranium Reconnaissance Evaluation program.

  6. Annual report on the state of RB reactor components and equipment, December 1997

    International Nuclear Information System (INIS)

    Milosevic, M.

    1997-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1997 the reactor lattice was not changed due to application of the coupled fast-thermal core HERBE. Total reactor operation amounted to 69.5 Wh with 66 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment, plan for forming new HERBE core, plan for regular annual maintenance of the reactor, and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  7. Causes of extended shutdown state of 'RA' research reactor in Vinca Institute

    International Nuclear Information System (INIS)

    Pesic, M.; Kolundzija, V.; Ljubenov, V.; Cupac, S.

    2001-01-01

    This paper describes the causes and reasons for extended shutdown state of RA research reactor in the 'Vinca' Institute of Nuclear Sciences. Technical and legal matters that led to decision to stop RA reactor operation in 1984 and further problems related to maintenance and preparation for continuation of operation are given. Influence of nuclear policy of Yugoslav government and the 'Vinca' Institute at prolongation of the reactor shutdown state, as consequence of changing of nuclear programme in the country and the world are discussed and underlined. An overview of the legislation in the field of nuclear safety and regulatory control of radiation sources and radioactive materials in Yugoslavia is presented. (author)

  8. Advance reactor and fuel-cycle systems--potentials and limitations for United States utilities

    International Nuclear Information System (INIS)

    Zebroski, E.L.; Williams, R.F.

    1979-01-01

    This paper reviews the potential benefits and limitations of advance reactor and fuel-cycle systems for United States utilities. The results of the review of advanced technologies show that for the near and midterm, the only advance reactor and fuel-cycle system with significant potential for United States utilities is the current LWR, and evolutionary, not revolutionary, enhancements. For the long term, the liquid-metal breeder reactor continues to be the most promising advance nuclear option. The major factors leading to this conclusion are summarized

  9. Current state of emergency preparedness at US power reactors

    International Nuclear Information System (INIS)

    Martin, J.B.; Grimes, B.K.

    1984-05-01

    The objectives were to ascertain whether the reactor operators were in compliance with NRC regulations and to determine whether they were taking appropriate actions to protect nuclear materials and facilities, the environment, and the health and safety of the public. The NRC has conducted this program with technical assistance from Pacific Northwest Laboratory. 5 references

  10. Current state of research on pressurized water reactor safety

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

    2018-03-01

    For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

  11. Prevalence and Significance of Parasites of Horses in Some States of Northern Nigeria

    Science.gov (United States)

    EHIZIBOLO, David O.; KAMANI, Joshua; EHIZIBOLO, Peter O.; EGWU, Kinsley O.; DOGO, Goni I.; SALAMI-SHINABA, Josiah O.

    2012-01-01

    This study was conducted to determine the prevalence and significance of parasites of horses in northern Nigeria. Blood and faecal samples were randomly collected from 243 horses from different stables in some states of northern Nigeria for laboratory analyses. Fifty-seven horses (23.5%) were found infected with parasites. The hemoparasites detected, 21 (8.6%), include Theileria equi, Babesia caballi, Trypanosoma vivax and Trypanosoma evansi. The endoparasites encountered, 29 (11.9%) were Strongylus spp., Strongyloides spp., Oxyuris equi, Parascaris equorum, Paragonimus spp. and Dicrocoelium spp., 3 (1.2%) was Eimeria spp. Four horses (1.6%) had mixed infection of hemo- and endoparasites. This preliminary finding shows that parasitism is a problem in the horse stables examined, and calls for proper stable hygiene, routine tick control and regular deworming programme. PMID:24833991

  12. Mixed enrichment core design for the NC State University PULSTAR Reactor

    International Nuclear Information System (INIS)

    Mayo, C.W.; Verghese, K.; Huo, Y.G.

    1997-12-01

    The North Carolina State University PULSTAR Reactor license was renewed for an additional 20 years of operation on April 30, 1997. The relicensing period added additional years to the facility operating time through the end of the second license period, increasing the excess reactivity needs as projected in 1988. In 1995, the Nuclear Reactor Program developed a strategic plan that addressed the future maintenance, development, and utilization of the facility. Goals resulting from this plan included increased academic utilization of the facility in accordance with its role as a university research facility, and increased industrial service use in accordance with the mission of a land grant university. The strategic plan was accepted, and it is the intent of the College of Engineering to operate the PULSTAR Reactor as a going concern through at least the end of the current license period. In order to reach the next relicensing review without prejudice due to low excess reactivity, it is desired to maintain sufficient excess reactivity so that, if relicensed again, the facility could continue to operate without affecting users until new fuel assistance was provided. During the NC State University license renewal, the operation of the PULSTAR Reactor at the State University of New York at Buffalo (SUNY Buffalo) was terminated. At that time, the SUNY Buffalo facility had about 240 unused PULSTAR Reactor fuel pins with 6% enrichment. The objective of the work reported here was to develop a mixed enrichment core design for the NC State University PULSTAR reactor which would: (1) demonstrate that 6% enriched SUNY buffalo fuel could be used in the NC State University PULSTAR Reactor within the existing technical specification safety limits for core physics parameters; (2) show that use of this fuel could permit operating the NC State University PULSTAR Reactor to 2017 with increased utilization; and (3) assure that the decision whether or not to relicense the facility would

  13. Present state of the liner of the reactor; Estado actual del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F; Raya A, R; Mazon R, R [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    When being presented to work the operation personnel of the reactor, on Monday January 10, 1983, they noticed that the reactor pool was overflowing of water and the floor of the room was partially flooded. The personnel proceeded to revise the feedwater systems to the pool, the Emergency Cooling System of the core and that of Water of Reinstatement, was found that the passing valve of this last it was lightly open. It was discovered that the water that was flooded in the floor of the room it came from the relief valves of the ports TW-1 and RW-2 and of three glides that were in the Thermal Column area. It was proceeded to lower the one level of water of the pool to their normal position and it was clean the water flooded in the salts. (Author)

  14. Data Validation Package April 2016 Groundwater and Surface Water Sampling at the Monticello, Utah, Disposal and Processing Sites August 2016

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [USDOE Office of Legacy Management, Washington, DC (United States); Smith, Fred [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-08-01

    This semiannual event includes sampling groundwater and surface water at the Monticello Disposal and Processing Sites. Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated) and Program Directive MNT-2016-01. Complete sample sets were collected from 42 of 48 planned locations (9 of 9 former mill site wells, 13 of 13 downgradient wells, 7 of 9 downgradient permeable reactive barrier wells, 4 of 7 seeps and wetlands, and 9 of 10 surface water locations). Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Locations R6-M3, SW00-01, Seep 1, Seep 2, and Seep 5 were not sampled due to insufficient water availability. A partial sample was collected at location R4-M3 due to insufficient water. All samples from the permeable reactive barrier wells were filtered as specified in the program directive. Duplicate samples were collected from surface water location Sorenson and from monitoring wells 92-07 and RlO-Ml. Water levels were measured at all sampled wells and an additional set of wells. See Attachment2, Trip Report for additional details. The contaminants of concern (COCs) for the Monticello sites are arsenic, manganese, molybdenum, nitrate+ nitrite as nitrogen (nitrate+ nitrite as N), selenium, uranium, and vanadium. Locations with COCs that exceeded remediation goals are listed in Table 1 and Table 2. Time-concentration graphs of the COCs for all groundwater and surface water locations are included in Attachment 3, Data Presentation. An assessment of anomalous data is included in Attachment 4.

  15. State-building, migration and economic development on the frontiers of northern Afghanistan and southern Tajikistan

    Directory of Open Access Journals (Sweden)

    Christian Bleuer

    2012-01-01

    Full Text Available The Kunduz River Valley of northern Afghanistan and the Vakhsh River Valley of southern Tajikistan followed what initially appear to be vastly different trajectories. Despite these two adjacent areas having had much in common throughout many periods of history, the present-day region of northern Afghanistan was eventually taken under the control of the Afghan state while the areas north of the Amu Darya and Panj River were to become part of the Soviet Union. However, instead of a divergent course of development and state-building, these two regions were subjected to very similar patterns of agricultural development and migration policies. “Empty” areas were to be populated, by force if necessary, wetlands were to be drained for agriculture, and cotton farming was to become pre-eminent. The end result in both areas was the creation of a socially diverse and economically significant region that was fully integrated into the modern state’s economy and politics. This article analyzes and compares the motives and implementation of the state-building projects in both of these now domestically important regions and finds remarkable similarities despite the obvious differences in the structure of the Afghan and Soviet states.

  16. State of the Art Report for Development of Control Element Drive Mechanism of the APR+ Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Seon; Choi, Suhn; Song, Chul Hwa

    2008-10-15

    Recently newly-developed nuclear reactors with increased safety and enhanced performance by developed countries in the nuclear area are competing in the global nuclear market. Several reactors, for example AP600 and AP1000 by Westinghouse Electric Co. in USA, EPR by Areva in Europe, APWR by Mitsubishi Heavy Industry in Japan in the pressurized power reactor, are competing to preoccupy the nuclear market during Nuclear Renaissance. Dedicated control element drive mechanism with enhanced performance and increased safety are developed for these new reactors. And load follow capability is required, and it is estimated that load follow requirement make design requirement of a control element drive mechanism harsh. It is necessary to review the current technical state of a control element drive mechanism. This work is aimed to review the design characteristics of a past and current control element drive mechanism for a nuclear reactor and to check the direction and goal of CEDM design development recently.

  17. State of the Art Report for Development of Control Element Drive Mechanism of the APR+ Reactor

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Choi, Suhn; Song, Chul Hwa

    2008-10-01

    Recently newly-developed nuclear reactors with increased safety and enhanced performance by developed countries in the nuclear area are competing in the global nuclear market. Several reactors, for example AP600 and AP1000 by Westinghouse Electric Co. in USA, EPR by Areva in Europe, APWR by Mitsubishi Heavy Industry in Japan in the pressurized power reactor, are competing to preoccupy the nuclear market during Nuclear Renaissance. Dedicated control element drive mechanism with enhanced performance and increased safety are developed for these new reactors. And load follow capability is required, and it is estimated that load follow requirement make design requirement of a control element drive mechanism harsh. It is necessary to review the current technical state of a control element drive mechanism. This work is aimed to review the design characteristics of a past and current control element drive mechanism for a nuclear reactor and to check the direction and goal of CEDM design development recently

  18. The ISRN has stated on the CABRI reactor restarting

    International Nuclear Information System (INIS)

    2009-01-01

    This paper presents the different issues examined by the ISRN (the French Institute of Radioprotection and Nuclear Safety) for the restarting of the pool type research CABRI reactor which is briefly described in appendix. These issues are: the design, realisation and monitoring of the new pressurised water test loop, the reassessment of the protection system limiting the reactivity injection during tests, inspection of fuel pencil condition, reassessment of safety studies, inspection of the condition of existing equipment which are essential for safety, reassessment of the seismic risk and of the fire risk, reassessment of operation conditions (personal radioprotection, human and organisational factors). An appendix contains the report by the Permanent Group of Experts for Nuclear Reactors with its recommendations

  19. Nuclear fuel for VVER reactors. Actual state and trends

    International Nuclear Information System (INIS)

    Molchanov, V.

    2011-01-01

    The main tasks concerning development of FA design, development and modernization of structural materials, improvement of technology of structural materials manufacturing and FA fabrication and development of methods and codes are discussed in this paper. The main features and expected benefit of implementation of second generation and third generation fuel assembly for VVER-440 Nuclear Fuel are given. A brief review of VVER-440 and VVER-1000 Nuclear Fuel development before 1997 since 2010 is shown. A summary of VVER-440 and VVER-1000 Nuclear Fuel Today, including details about TVSA-PLUS, TVSA-ALFA, TVSA-12 and NPP-2006 Phase 2 tasks (2010-2012) is presented. In conclusion, as a result of large scope of R and D performed by leading enterprises of nuclear industry modern nuclear fuel for VVER reactors is developed, implemented and successfully operated. Fuel performance (burnup, lifetime, fuel cycles, operating reliability, etc.) meets the level of world's producers of nuclear fuel for commercial reactors

  20. Present state of the liner of the reactor

    International Nuclear Information System (INIS)

    Aguilar H, F.; Raya A, R.; Mazon R, R.

    2001-07-01

    When being presented to work the operation personnel of the reactor, on Monday January 10, 1983, they noticed that the reactor pool was overflowing of water and the floor of the room was partially flooded. The personnel proceeded to revise the feedwater systems to the pool, the Emergency Cooling System of the core and that of Water of Reinstatement, was found that the passing valve of this last it was lightly open. It was discovered that the water that was flooded in the floor of the room it came from the relief valves of the ports TW-1 and RW-2 and of three glides that were in the Thermal Column area. It was proceeded to lower the one level of water of the pool to their normal position and it was clean the water flooded in the salts. (Author)

  1. Rayleigh and Love Wave Phase Velocities in the Northern Gulf Coast of the United States

    Science.gov (United States)

    Li, A.; Yao, Y.

    2017-12-01

    The last major tectonic event in the northern Gulf Coast of the United States is Mesozoic continental rifting that formed the Gulf of Mexico. This area also experienced igneous activity and local uplifts during Cretaceous. To investigate lithosphere evolution associated with the rifting and igneous activity, we construct Rayleigh and Love wave phase velocity models at the periods of 6 s to 125 s in the northern Gulf Coast from Louisiana to Alabama including the eastern Ouachita and southern Appalachian orogeny. The phase velocities are derived from ambient noise and earthquake data recorded at the 120 USArray Transportable Array stations. At periods below 20 s, phase velocity maps are characterized by significant low velocities in the Interior Salt Basin and Gulf Coast Basin, reflecting the effects of thick sediments. The northern Louisiana and southern Arkansas are imaged as a low velocity anomaly in Rayleigh wave models but a high velocity anomaly of Love wave at the periods of 14 s to 30 s, indicating strong lower crust extension to the Ouachita front. High velocity is present in the Mississippi Valley Graben from period 20 s to 35 s, probably reflecting a thin crust or high-velocity lower crust. At longer periods, low velocities are along the Mississippi River to the Gulf Coast Basin, and high velocity anomaly mainly locates in the Black Warrior Basin between the Ouachita Belt and Appalachian Orogeny. The magnitude of anomalies in Love wave images is much smaller than that in Rayleigh wave models, which is probably due to radial anisotropy in the upper mantle. A 3-D anisotropic shear velocity model will be developed from the phase velocities and will provide more details for the crust and upper mantle structure beneath the northern Gulf of Mexico continental margin.

  2. Use of the Oregon State University TRIGA reactor for education and training

    International Nuclear Information System (INIS)

    Dodd, B.

    1989-01-01

    This paper summarizes the recent use of the Oregon State University TRIGA Reactor (OSTR) for education and training. In particular, data covering the last 5 yr are presented, which cover education through formal university classes, theses, public information, and school programs. Training is covered by presenting data on domestic and foreign reactor operator training, health physics training, and neutron activation analysis training. While education and training only occupy ∼16% of the OSTR's total use time, nevertheless, this is an important mission of all nonpower reactors that cannot be performed effectively in any other way

  3. Status of fast breeder reactor development in the United States of America

    International Nuclear Information System (INIS)

    Horton, K.E.

    1983-01-01

    The goal of the United States Liquid Metal Fast Breeder Reactor (LMFBR) program is to develop the technology to the point that the private sector can deploy a safe, economic breeder reactor. The LMFBR will provide virtually inexhaustible supplies of electrical energy for the long term and will provide additional confidence to LWR nuclear deployment in the near term. The LMFBR program consists of a streamlined research and development effort focussing on those actions needed to enable private sector financing of industrial deployment including plant demonstration and technology efforts in reactor fuels, components, materials, physics, and safety

  4. Effect of different materials in the performance of solar reactors deployed in Jaiba, Minas Gerais state

    Energy Technology Data Exchange (ETDEWEB)

    Sartori, Marcia Aparecida; Soares, Antonio Alves; Soares, Adilson Rodrigues; Batista, Rafael Oliveira; Leite, Caio Vinicius [Universidade Federal de Vicosa (DEA/UFV), MG (Brazil). Dept. de Engenharia Agricola

    2008-07-01

    This study aimed to analyze the effect of different materials (masonry, butyl canvas and fiberglass) in the performance of solar reactors deployed in the city of Jaiba, Minas Gerais State. To do so, mini-stations to treat the domestic sewage were assembled. During the tests, samples of the effluent were collected upstream and downstream of the septic tank and the solar reactor. Fecal coliforms, BOD and COD were quantified in laboratory. The results indicated that the materials tested for construction of the reactor did not influence the solar disinfection of fecal coliforms. (author)

  5. Status of development and licensing support for advanced liquid metal reactors in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, D R [Argonne National Laboratory, Argonne, IL (United States); Gyorey, G [General Electric, San Jose, CA (United States)

    1991-07-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  6. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment

  7. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  8. State of oil pollution in the northern Arabian Sea after the 1991 Gulf oil spill

    Digital Repository Service at National Institute of Oceanography (India)

    Sengupta, R.; Fondekar, S.P.; Alagarsamy, R.

    stream_size 30182 stream_content_type text/plain stream_name Mar_Pollut_Bull_27_85.pdf.txt stream_source_info Mar_Pollut_Bull_27_85.pdf.txt Content-Encoding UTF-8 Content-Type text/plain; charset=UTF-8 Marine Pollution... Bulletin, Volume 27, pp. 85-91, 1993. 0025-326X/93 $6.00+0.00 Printed in Great Britain. O 1993 Pergamon Press Ltd State of Oil Pollution in the Northern Arabian Sea after the 1991 Gulf Oil Spill R. SEN GUPTA, S. P. FONDEKAR and R. ALAGARSAMY National...

  9. Hedychium putaoense (Zingiberaceae, a new species from Putao, Kachin State, Northern Myanmar

    Directory of Open Access Journals (Sweden)

    Hong-Bo Ding

    2018-01-01

    Full Text Available Hedychium putaoense Y.H. Tan & H.B. Ding, a new species of Zingiberaceae from Putao, Kachin state, Northern Myanmar, is described and illustrated. It is similar to H. densiflorum Wall. and H. longipedunculatum A.R.K. Sastry & D.M. Verma, but differs by its very small bract (4–6 × 2.5–3 mm vs. 18–19 × 5–5.5 mm and ca. 11 × 7 mm, respectively, semicircle and dark red bracteole, orange flower and broadly falcate to lanceolate lateral staminodes.

  10. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    International Nuclear Information System (INIS)

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

  11. State of the art on reactor noise analysis

    International Nuclear Information System (INIS)

    Bernard, P.; Fry, D.; Stegemann, D.; Van Dam, H.

    1986-01-01

    This report is the result of the work of a task force sponsored by the NEA Committee on Reactor Physics (NEACRP) and is divided into six chapters: 1. Loose-Parts Detection and Acoustic Monitoring, 2. Thermal Hydraulics Surveillance, 3. Flow Measurements, 4. Vibration Monitoring, 5. Surveillance Systems and Evaluation Methods, and 6. System Dynamic Analysis. Each chapter summarizes the current situation in noise analysis techniques with emphasis on the following aspects: . physical quantities considered, . possible anomalies involved, . sensors used for the detection, and . conditions of applications. The remainder of each chapter discusses future trends and recommendations

  12. Acoustic monitoring of the BOR-60 reactor circulating pump state

    International Nuclear Information System (INIS)

    Efimov, V.N.; Myntsov, A.A.

    1988-01-01

    Diagnostics methods for circulation pumps of the experimental BOR-60 fast reactor are described. The results of signal processing during a microcompain, as well as detected anomalies in pump operation in the earth stage are presented. Analysis carried out for an acoustic signal envelope has shown high efficiency of the method. When oscillations of a mechanical shaft are present, the envelope level increases 1.5 times. More detailed investigation is carried out by the analysis of the spectrum of the pump acoustic signal envelope. During abnormal operation there are peaks, corresponding to the circulation frequency, and harmonics multiple of it, in the spectrum. 6 figs

  13. The Text of the Agreement for the Application of Agency Safeguards to United States Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-08-14

    The text of the Agreement between the Agency and the Government of the United States of America for the application of Agency safeguards to United States reactor facilities, which was signed on 15 June 1964 and entered into force on 1 August 1964, is reproduced in this document for the information of all Members.

  14. Characterization of the TRIGA Mark II reactor full-power steady state

    Energy Technology Data Exchange (ETDEWEB)

    Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

    2016-04-15

    Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

  15. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  16. Phytogeographical patterns of dry forests sensu stricto in northern Minas Gerais State, Brazil.

    Science.gov (United States)

    Arruda, Daniel M; Ferreira-Júnior, Walnir G; Duque-Brasil, Reinaldo; Schaefer, Carlos E R

    2013-01-01

    The Deciduous Complex that occurs in northern Minas Gerais State, Brazil, raises questions about the floristic affinities of these formations in relation to neighboring phytogeographical domains. Little is known about the identity of the seasonal forest formations that comprise this complex, or about its relationships to abiotic components, such as soils, topography and climate. This study aimed to recognize the patterns of floristic similarity of all studied fragments of dry forest of northern Minas Gerais with soil and climate attributes, based on the available database. Cluster analysis indicated the existence of two floristic groups that had clear associations with either the Koppen's BSh (semi-arid) or Aw (seasonal tropical) climates. Likewise, the subdivisions of these groups showed clear associations with the dominant soil classes in the region. The Red-Yellow Latosol is the dominant soil classes in the BSh climatic domain, seconded by alluvial areas associated with Fluvic Neosols. The Aw domain comprised a much varied set of soils: Nitosols, Argisols, Cambisols and Litholic Neosols, most derived from the Bambuí limestone/slate formation. The ecotonal nature of northern Minas Gerais State provides a complex interaction between the flora of neighboring phytogeographical domains. This, allied to pedogeomorphological factors, allowed a better understanding of the effects of late Quaternary climate changes for the Deciduous Complex evolution. We conclude that the Latosols under present-day semi-arid climates (BSh) are relicts of former wetter climates, during which humid forest (semideciduous) expansion took place. Later, these semideciduous forests were subjected to a much drier climate, when selection for deciduousness led to the present-days Deciduous Complex scenario.

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

    Directory of Open Access Journals (Sweden)

    Subagjo

    2014-07-01

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

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

    International Nuclear Information System (INIS)

    Mayo, C.W.

    1992-01-01

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

  19. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1991-01-01

    An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Subsequent to the directive to shut down the Fast Flux Test Facility in early 1990, a variety of activities were undertaken to provide support for continued operation. The United States has made substantial progress in achieving ALMR program objectives. The metal fuel cycle is designed to recycle and burn its own actiniums, and has the potential to be a very effective burner of actiniums generated in the LWRs. The current emphasis in the IFR Program is on the comprehensive development of the IFR (Integral Fast Reactor) technology, to be followed by a period of technology demonstration which would verify the economic feasibility of the concept. The United States has been active in international cooperative activities in the fast reactor sector since 1969. (author). 11 figs, 1 tab

  20. Optimization and control of a novel upflow anaerobic solid-state (UASS) reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mumme, J.; Linke, B. [Leibniz Inst. for Agricultural Engineering, Potsdam (Germany); Tolle, R. [Humboldt Univ., Berlin (Germany). Dept. of Biosystems Technology

    2010-07-01

    Optimization and control strategies for a newly developed upflow anaerobic solid-state (UASS) reactor equipped with liquor recirculation were investigated. The UASS reactor converts solid biomass into biogas while the particulate organic matter (POM) ascends in the form of a solid-state bed (SSB) driven by the adherence of self-produced micro gas bubbles. Performance data and technical characteristics were obtained from a technical scale semi-automatic 400 L UASS reactor that operated for 117 days with maize silage under thermophilic conditions at 55 degrees C. The process liquor was continuously recirculated through separate methanogenic reactors in order to prevent an accumulation of volatile fatty acids. Emphasis was placed on determining the gas and metabolite production. The volatile solids (VS) loading rate was fixed at 5 g per litre per day. The methane production rate of the UASS reactor stabilized between 1.5 and 2.0 L per litre per day. The average volatile solids (VS) methane yield of the maize silage was 380 L per kg. The liquor exchange was found to play an important role in the performance and stability of the digestion process. Although low exchange rates can cause process failure by acidification, high exchange rates have the risk of clogging inside the SSB. It was concluded that the UASS reactor is a viable solution for the digestion of various organic matter.

  1. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

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

  2. Current drive efficiency requirements for an attractive steady-state reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tonon, G

    1994-12-31

    The expected values of the figure of merit and the electrical efficiency of various non-inductive current drive methods are considered. The main experimental results achieved today with neutral beams and radiofrequency systems are summarized. Taking into account the simplified energy flow diagram of a steady state reactor, the figure of merit and the electrical efficiency values which are necessary in order to envisage an attractive steady-state reactor are determined. These values are compared to the theoretical predictions. (author). 16 refs., 11 figs., 2 tabs.

  3. Modelling dynamic processes in a nuclear reactor by state change modal method

    Science.gov (United States)

    Avvakumov, A. V.; Strizhov, V. F.; Vabishchevich, P. N.; Vasilev, A. O.

    2017-12-01

    Modelling of dynamic processes in nuclear reactors is carried out, mainly, using the multigroup neutron diffusion approximation. The basic model includes a multidimensional set of coupled parabolic equations and ordinary differential equations. Dynamic processes are modelled by a successive change of the reactor states. It is considered that the transition from one state to another occurs promptly. In the modal method the approximate solution is represented as eigenfunction expansion. The numerical-analytical method is based on the use of dominant time-eigenvalues of a group diffusion model taking into account delayed neutrons.

  4. Tabular equation of state of lithium for laser-fusion reactor studies

    International Nuclear Information System (INIS)

    Young, D.A.; Ross, M.; Rogers, F.J.

    1979-01-01

    A tabular lithium equation of state was formulated from three separate equation-of-state models to carry out hydrodynamic simulations of a lithium-waterfall laser-fusion reactor. The models we used are: ACTEX for the ionized fluid, soft-sphere for the liquid and vapor, and pseudopotential for the hot, dense liquid. The models are smoothly joined over the range of density and temperature conditions appropriate for a laser-fusion reactor. We also fitted the models into two forms suitable for hydrodynamic calculations

  5. Tabular equation of state of lithium for laser-fusion reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Young, D.A.; Ross, M.; Rogers, F.J.

    1979-01-19

    A tabular lithium equation of state was formulated from three separate equation-of-state models to carry out hydrodynamic simulations of a lithium-waterfall laser-fusion reactor. The models we used are: ACTEX for the ionized fluid, soft-sphere for the liquid and vapor, and pseudopotential for the hot, dense liquid. The models are smoothly joined over the range of density and temperature conditions appropriate for a laser-fusion reactor. We also fitted the models into two forms suitable for hydrodynamic calculations.

  6. Current drive efficiency requirements for an attractive steady-state reactor

    International Nuclear Information System (INIS)

    Tonon, G.

    1994-01-01

    The expected values of the figure of merit and the electrical efficiency of various non-inductive current drive methods are considered. The main experimental results achieved today with neutral beams and radiofrequency systems are summarized. Taking into account the simplified energy flow diagram of a steady state reactor, the figure of merit and the electrical efficiency values which are necessary in order to envisage an attractive steady-state reactor are determined. These values are compared to the theoretical predictions. (author). 16 refs., 11 figs., 2 tabs

  7. On achieving the state's household recycling target: A case study of Northern New Jersey, USA

    International Nuclear Information System (INIS)

    Otegbeye, M.; Abdel-Malek, L.; Hsieh, H.N.; Meegoda, J.N.

    2009-01-01

    In recent times, the State of New Jersey (USA) has been making attempts at promoting recycling as an environmentally friendly means of attaining self-sufficiency at waste disposal, and the state has put in place a 50% recycling target for its municipal solid waste stream. While the environmental benefits of recycling are obvious, a recycling program must be cost effective to ensure its long-term sustainability. In this paper, a linear programming model is developed to examine the current state of recycling in selected counties in Northern New Jersey and assess the needs to achieve the state's recycling goal in these areas. The optimum quantities of waste to be sent to the different waste facilities, which include landfills, incinerators, transfer stations, recycling and composting plants, are determined by the model. The study shows that for these counties, the gap between the current waste practices where the recycling rate stands at 32% and the state's goal can be bridged by more efficient utilization of existing facilities and reasonable investment in expanding those for recycling activities

  8. Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

  9. Employability of People with Disabilities in the Northern States of Peninsular Malaysia: Employers’ Perspective

    Directory of Open Access Journals (Sweden)

    Khoo Suet Leng

    2011-10-01

    Full Text Available Purpose: This study investigates employers’ perspective towards employing people with disabilities  in the northern states of Peninsular Malaysia. The research also endeavoured to identify factors that promote or hinder gainful employment of people with disabilities in Malaysia.Method: The data was collected through postal questionnaires distributed to several types of industries in the northern region of Malaysia.Results: The results indicated that most of the employers are in favour of employing persons with disabilities. However, very few have such enabling policies, or a mechanism to handle issues related to persons with disabilities, or a built environment which is fully accessible to persons with disabilities.  They are also concerned about the  ability of workers with disabilities to comprehend and follow orders, as well as the costs involved in employing and training them. These results imply that if employers  want to fulfil their intentions of recruiting persons with disabilities, a lot has to be done to employ and sustain them in their jobs.Key words: Employment, People with Disabilities, Malaysiadoi 10.5463/DCID.v22i2.28

  10. Echovirus 30 associated with cases of aseptic meningitis in state of Pará, Northern Brazil

    Directory of Open Access Journals (Sweden)

    Ceyla Maria Oeiras de Castro

    2009-05-01

    Full Text Available Investigation of the aetiology of viral meningitis in Brazil is most often restricted to cases that occur in the Southern and Southeastern Regions; therefore, the purpose of this study is to describe the viral meningitis cases that occurred in state of Pará, Northern Brazil, from January 2005-December 2006. The detection of enterovirus (EV in cerebrospinal fluid was performed using cell culture techniques, RT-PCR, nested PCR and nucleotide sequencing. The ages of the 91 patients ranged from 60 years old (median age 15.90 years. Fever (87.1%, headache (77.0%, vomiting (61.5% and stiffness (61.5% were the most frequent symptoms. Of 91 samples analyzed, 18 (19.8% were positive for EV. Twelve were detected only by RT- PCR followed by nested PCR, whereas six were found by both cell culture and RT-PCR. From the last group, five were sequenced and classified as echovirus 30 (Echo 30. Phylogenetic analyses revealed that Echo 30 detected in Northern Brazil clustered within a unique group with a bootstrap value of 100% and could constitute a new subgroup (4c according to the phylogenetic tree described by Oberste et al. (1999. This study described the first molecular characterization of Echo 30 in Brazil and this will certainly contribute to future molecular analyses involving strains detected in other regions of Brazil.

  11. Status of fusion technology development in JAERI stressing steady-state operation for future reactors

    International Nuclear Information System (INIS)

    Matsuda, Shinzaburo

    2000-01-01

    This paper reports on the progress of the fusion reactor technologies developed at the Japan Atomic Energy Research Institute (JAERI) and expected to lead to a future steady state operation reactor. In particular, superconducting coil technology for plasma confinement, NBI and RF systems technology for plasma control and current drive, fueling and pumping systems technology for particle control, heat removal technology, and development of long life materials are highlighted as the important key elements for the future steady state operation. It will be discussed how these key technologies have already been developed by the ITER (International Thermonuclear Experimental Reactor) technology R and D as well as by the Japanese domestic program, and which technologies are planned for the near future

  12. On the optimization of a steady-state bootstrap-reactor

    International Nuclear Information System (INIS)

    Polevoy, A.R.; Martynov, A.A.; Medvedev, S.Yu.

    1993-01-01

    A commercial fusion tokamak-reactor may be economically acceptable only for low recirculating power fraction r 0 ≡ P CD /P α BS ≡I BS /I > 0.9 to sustain the steady-state operation mode for high plasma densities > 1.5 10 20 m -3 , fulfilled the divertor conditions. This paper presents the approximate expressions for the optimal set of reactor parameters for r BS /I∼1, based on the self-consistent plasma simulations by 1.5D ASTRA code. The linear MHD stability analysis for ideal n=1 kink and ballooning modes has been carried out to determine the conditions of stabilization for bootstrap steady state tokamak reactor BSSTR configurations. (author) 10 refs., 1 tab

  13. Moving into the 21st century - The United States' Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Reilly, Jill E.

    1999-01-01

    Since 1996, when the United States Department of Energy and the Department of State jointly adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, twelve shipments totaling 2,985 MTR and TRIGA spent nuclear fuel assemblies from research reactors around the world have been accepted into the United States. These shipments have contained approximately 1.7 metric tons of HEU and 0.6 metric tons of LEU. Foreign research reactor operators played a significant role in this success. A new milestone in the acceptance program occurred during the summer of 1999 with the arrival of TRIGA spent nuclear fuel from Europe through the Charleston Naval Weapons Station via the Savannah River Site to the Idaho National Engineering and Environmental Laboratory. This shipment consisted of five casks of TRIGA spent nuclear fuel from research reactors in Germany, Italy, Slovenia, and Romania. These casks were transported by truck approximately 2,400 miles across the United States (one cask packaged in an ISO container per truck). Drawing upon lessons learned in previous shipments, significant technical, legal, and political challenges were addressed to complete this cross-country shipment. Other program activities since the last RERTR meeting have included: formulation of a methodology to determine the quantity of spent nuclear fuel in a damaged condition that may be transported in a particular cask (containment analysis for transportation casks); publication of clarification of the fee policy; and continued planning for the outyears of the acceptance policy including review of reactors and eligible material quantities. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues to demonstrate success due to the continuing commitment between the United States and the research reactor community to make this program work. We strongly encourage all eligible research reactors to decide as soon as possible to

  14. The sexual games of the body politic: fantasy and state violence in Northern Ireland.

    Science.gov (United States)

    Aretxaga, B

    2001-03-01

    This article analyzes the practice of strip searching women political prisoners in Northern Ireland as a violent technology of control aimed at breaking the political identity of prisoners. Focusing on a controversial case of a mass strip search carried out in 1992, the article examines the phantasmatic investements pervading this seemingly rational technology of control. Using a psychoanalytic notion of fantasy against the backdrop of a Foucaultian theory of power, this article argues that strip searches constitute a gendered form of political domination driven by, and performed within, a phantasmatic scenario of sexual violence. In this scenario both the political and gender identities of prisoners are re-inscribed with the power of a state acting as a male body politic. The article argues that the phantasmatic support of rational technologies of control betrays the contingent and shifting character of domination as well as its ambiguous effects.

  15. New steady-state microbial community compositions and process performances in biogas reactors induced by temperature disturbances

    DEFF Research Database (Denmark)

    Luo, Gang; De Francisci, Davide; Kougias, Panagiotis

    2015-01-01

    that stochastic factors had a minor role in shaping the profile of the microbial community composition and activity in biogas reactors. On the contrary, temperature disturbance was found to play an important role in the microbial community composition as well as process performance for biogas reactors. Although...... three different temperature disturbances were applied to each biogas reactor, the increased methane yields (around 10% higher) and decreased volatile fatty acids (VFAs) concentrations at steady state were found in all three reactors after the temperature disturbances. After the temperature disturbance...... in shaping the profile of the microbial community composition and activity in biogas reactors. New steady-state microbial community profiles and reactor performances were observed in all the biogas reactors after the temperature disturbance....

  16. Compilation of data and descriptions for United States and foreign liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Appleby, E.R.

    1975-08-01

    This document is a compilation of design and engineering information pertaining to liquid metal cooled fast breeder reactors which have operated, are operating, or are currently under construction, in the United States and abroad. All data has been taken from publicly available documents, journals, and books

  17. Nuclear reactors built, being built, or planned in the Unites States as of June 30, 1981

    International Nuclear Information System (INIS)

    Goulden, A.M.

    1983-01-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of June 30, 1981, which are capable of sustaining a nuclear chain reaction. Information is presented in five parts, each of which is categorized by primary function or pupose: civilian, military, production, export, and critical assembly facilities

  18. Nuclear reactors built, being built, or planned in the United States as of December 31, 1980

    International Nuclear Information System (INIS)

    1981-04-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1980, which are capable of sustaining a nuclear chain reaction. Information is presented in five parts, each of which is categorized by primary function or purpose: civilian, military, production, export, and critical assembly facilities

  19. Education and research at the Ohio State University nuclear reactor laboratory

    International Nuclear Information System (INIS)

    Miller, D.W.; Myser, R.D.; Talnagi, J.W.

    1989-01-01

    The educational and research activities at the Ohio State University Nuclear Reactor Laboratory (OSUNRL) are discussed in this paper. A brief description of an OSUNRL facility improvement program and its expected impact on research is presented. The overall long-term goal of the OSUNRL is to support the comprehensive education, research, and service mission of OSU

  20. Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

    International Nuclear Information System (INIS)

    Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

    2010-01-01

    In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

  1. Power reactor services provided by the Penn State Radiation Science and Engineering Center

    International Nuclear Information System (INIS)

    Voth, M.H.; Jester, W.A.

    1993-01-01

    The power reactor industry emerged from extensive research and development performed at nonpower reactors (NPRs). As the industry matures, NPRs continue to support and enhance power reactor technology. With the closure of many government and private industry NPRS, there is an increasing call for the 33 universities with operating research reactors to provide the needed services. The Penn State Radiation Science and Engineering Center (RSEC) includes a 1-MW pool-type pulsing TRIGA reactor, a neutron beam laboratory with real-time neutron radiography equipment, hot cells with master-slave manipulators for remote handling of radioactive materials, a gamma-ray irradiation pool, a low-level radiation monitoring laboratory, and extensive equipment for radiation monitoring, dosimetry, and material properties determination. While equipment is heavily utilized in the instructional and academic research programs, significant time remains available for service work. Cost recovery for service work generates income for personnel, equipment maintenance, and facility improvements. With decreasing federal and state funding for educational programs, it is increasingly important that facilities be fully utilized to generate supplementary revenue. The following are examples of such work performed at the RSEC

  2. Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

    International Nuclear Information System (INIS)

    Pryor, R.J.; Maloney, K.J.

    1990-10-01

    This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs

  3. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

    International Nuclear Information System (INIS)

    Schickler, R.A.; Marcum, W.R.; Reese, S.R.

    2013-01-01

    Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

  4. Burn cycle requirements comparison of pulsed and steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ehst, D.A.

    1983-12-01

    Burn cycle parameters and energy transfer system requirements were analyzed for an 8-m commercial tokamak reactor using four types of cycles: conventional, hybrid, internal transformer, and steady state. Not surprisingly, steady state is the best burn mode if it can be achieved. The hybrid cycle is a promising alternative to the conventional. In contrast, the internal transformer cycle does not appear attractive for the size tokamak in question

  5. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1990-01-01

    The United States have made substantial progress in achieving Advanced Liquid Metal Reactor (ALMR) program objectives. A decision was made in 1988 to select the General Electric ALMR concept known as PRISM (Power Reactor Innovative Safe Module) for advanced conceptual design. A 3-year contract was awarded to General Electric in January of last year for concentrated trade-off studies and advanced design development. The strategy is to integrate those advancements that best meet program objectives into a national ALMR system concept. (author). 10 figs, 1 tab

  6. A method for statistical steady state thermal analysis of reactor cores

    International Nuclear Information System (INIS)

    Whetton, P.A.

    1981-01-01

    In a previous publication the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analysis are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations. (orig.)

  7. The Northern Climate Exchange Gap Analysis Project : an assessment of the current state of knowledge about the impacts of climate change in northern Canada

    International Nuclear Information System (INIS)

    2002-01-01

    The Northern Climate ExChange (NCE) Gap Analysis Project was launched in 1999 with an objective to assess the state of knowledge on climate change in northern Canada. Resulting products of the project have included the Infosource Database, an on-line database of published climate change research related to the Canadian North, the Directory of Contacts, another on-line database of interested parties to climate change issues, and a set of tables that rate the level of available information on climate change as it relates to natural, economic and community systems. Other products include a report of a workshop on climate change research, 2 reports assessing the level of traditional northern knowledge about climate change, 2 reports assessing the completeness and value of the Infosource Database, a web site for NCE, and this report. All products are available to the public on the Internet or on a CD-ROM. The NCE Gap Analysis Project has shown there are inequalities in the amount of information across different systems, and that there is more knowledge on predicted temperature changes than for other climate components. The study notes that there are strong regional trends for compiled knowledge, with some regions having been better studied than others. The project revealed that traditional knowledge of climate change has not been well documented, and that more information exists about climate change impacts on biological systems with an economic component than those without economic significance. refs., tabs., figs

  8. Progress of the United States foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, D.G.; Clapper, M.; Thrower, A.W.

    2002-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program has completed 23 shipments. Almost 5000 spent fuel assemblies from eligible research reactors throughout the world have been accepted into the United States under this program. Over the past year, another cross-country shipment of fuel was accomplished, as well as two additional shipments in the fourth quarter of calendar year 2001. These shipments attracted considerable safeguards oversight since they occurred post September 11. Recent guidance from the Nuclear Regulatory Commission (NRC) pertaining to security and safeguards issues deals directly with the transport of nuclear material. Since the Acceptance Program has consistently applied above regulatory safety enhancements in transport of spent nuclear fuel, this guidance did not adversely effect the Program. As the Program draws closer to its termination date, an increased number of requests for program extension are received. Currently, there are no plans to extend the policy beyond its current expiration date; therefore, eligible reactor operators interested in participating in this program are strongly encouraged to evaluate their inventory and plan for future shipments as soon as possible. (author)

  9. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  10. Grassland to shrubland state transitions enhance carbon sequestration in the northern Chihuahuan Desert.

    Science.gov (United States)

    Petrie, M D; Collins, S L; Swann, A M; Ford, P L; Litvak, M E

    2015-03-01

    The replacement of native C4 -dominated grassland by C3 -dominated shrubland is considered an ecological state transition where different ecological communities can exist under similar environmental conditions. These state transitions are occurring globally, and may be exacerbated by climate change. One consequence of the global increase in woody vegetation may be enhanced ecosystem carbon sequestration, although the responses of arid and semiarid ecosystems may be highly variable. During a drier than average period from 2007 to 2011 in the northern Chihuahuan Desert, we found established shrubland to sequester 49 g C m(-2) yr(-1) on average, while nearby native C4 grassland was a net source of 31 g C m(-2) yr(-1) over this same period. Differences in C exchange between these ecosystems were pronounced--grassland had similar productivity compared to shrubland but experienced higher C efflux via ecosystem respiration, while shrubland was a consistent C sink because of a longer growing season and lower ecosystem respiration. At daily timescales, rates of carbon exchange were more sensitive to soil moisture variation in grassland than shrubland, such that grassland had a net uptake of C when wet but lost C when dry. Thus, even under unfavorable, drier than average climate conditions, the state transition from grassland to shrubland resulted in a substantial increase in terrestrial C sequestration. These results illustrate the inherent tradeoffs in quantifying ecosystem services that result from ecological state transitions, such as shrub encroachment. In this case, the deleterious changes to ecosystem services often linked to grassland to shrubland state transitions may at least be partially offset by increased ecosystem carbon sequestration. © 2014 John Wiley & Sons Ltd.

  11. The United States fluoride-salt-cooled high-temperature reactor program

    International Nuclear Information System (INIS)

    Holcomb, David E.

    2013-01-01

    The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid-metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High-temperature gas-cooled reactors provide experience with coated-particle fuel and graphite components. Light-water reactors show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

  12. Time-series Oxygen-18 Precipitation Isoscapes for Canada and the Northern United States

    Science.gov (United States)

    Delavau, Carly J.; Chun, Kwok P.; Stadnyk, Tricia A.; Birks, S. Jean; Welker, Jeffrey M.

    2014-05-01

    The present and past hydrological cycle from the watershed to regional scale can be greatly enhanced using water isotopes (δ18O and δ2H), displayed today as isoscapes. The development of water isoscapes has both hydrological and ecological applications, such as ground water recharge and food web ecology, and can provide critical information when observations are not available due to spatial and temporal gaps in sampling and data networks. This study focuses on the creation of δ18O precipitation (δ18Oppt) isoscapes at a monthly temporal frequency across Canada and the northern United States (US) utilizing CNIP (Canadian Network for Isotopes in Precipitation) and USNIP (United States Network for Isotopes in Precipitation) measurements. Multiple linear stepwise regressions of CNIP and USNIP observations alongside NARR (North American Regional Reanalysis) climatological variables, teleconnection indices, and geographic indicators are utilized to create empirical models that predict the δ18O of monthly precipitation across Canada and the northern US. Pooling information from nearby locations within a region can be useful due to the similarity of processes and mechanisms controlling the variability of δ18O. We expect similarity in the controls on isotopic composition to strengthen the correlation between δ18Oppt and predictor variables, resulting in model simulation improvements. For this reason, three different regionalization approaches are used to separate the study domain into 'isotope zones' to explore the effect of regionalization on model performance. This methodology results in 15 empirical models, five within each regionalization. A split sample calibration and validation approach is employed for model development, and parameter selection is based on demonstrated improvement of the Akaike Information Criteria (AIC). Simulation results indicate the empirical models are generally able to capture the overall monthly variability in δ18Oppt. For the three

  13. Summary of the fourth conference on United States utility experience in reactor noise analysis

    International Nuclear Information System (INIS)

    Fry, D.N.

    1987-01-01

    The fourth informal conference on United States utility experience in reactor noise analysis and loose-part monitoring was held at the Northeast Utilities Service Company offices in Hartford, Connecticut, May 12-14, 1987. Host and general chairman for the meeting was J.V. Persio of Northeast Utilities. This conference provided a forum where utilities could share information on reactor noise analysis on an informal basis. There were about 60 attendees at the meeting representing 10 utilities, 3 reactor vendors, 8 consulting organizations, and 4 universities and research laboratories. Twenty-three papers were presented at the conference, dealing with various aspects of loose-part monitoring, neutron noise analysis, and standards activities

  14. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

  15. Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

    International Nuclear Information System (INIS)

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

    1983-07-01

    The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

  16. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  17. Monticello Nuclear Generating Plant, Unit 1. Semiannual operating report, July--December 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 1,103,754 MWH(e) with the reactor on line 2,743 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, and environmental radiation monitoring

  18. Monticello Nuclear Generating Plant, Unit 1. Semiannual operating report No. 9, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    Net electrical power generated was 1,775,704 MWH with the reactor on line 3,583 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, abnormal occurrences, and radiation environmental monitoring. (FS)

  19. Decommissioning of research reactors: Evolution, state of the art, open issues

    International Nuclear Information System (INIS)

    2006-01-01

    Many research reactors throughout the world date from the original nuclear research programmes in Member States. Consequently, dozens of old research reactors are candidates for near term decommissioning in parallel with progressive ageing and technical and economic obsolescence. Many of them are located in countries/institutions that, although familiar with the operation and management of their reactors, do not necessarily have adequate expertise and technologies for planning and implementing state of the art decommissioning projects. It is felt that IAEA reports may contribute to the awareness of technologies and know-how already tested successfully elsewhere. This report addresses a subject area that was dealt with earlier by two IAEA publications, namely, Planning and Management for the Decommissioning of Research Reactors and Other Small Nuclear Facilities (Technical Reports Series No. 351) and Decommissioning Techniques for Research Reactors (Technical Reports Series No. 373). This publication updates those reports in view of the technological progress, experience gained and the progressive ageing of research reactors, many of which have already reached the permanent shutdown stage and should be decommissioned soon. It is intended to contribute to the systematic coverage of the entire range of activities that have been addressed by the IAEA's decommissioning work in past years. The perspective of the report is historical, in that relevant issues are identified as solved, pending, or emerging. Much of the information provided in this report will also be of use for the decommissioning of nuclear power plants and other nuclear facilities. A Technical Committee Meeting on this subject was held in Vienna from 17 to 21 May 2004, at which the participants reviewed a draft report written by consultants from Canada, Germany, Israel, the Russian Federation and the United Kingdom

  20. The United States foreign research reactor spent nuclear fuel acceptance program: Proposal to modify the program

    International Nuclear Information System (INIS)

    Messick, C.E.

    2005-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. The policy was slated to expire in May 2009. However, in October 2003, a petition requesting a program extension was delivered to the United States Secretary of Energy from a group of research reactor operators from foreign countries. In April 2004, the Secretary directed DOE undertake an analysis, as required by the National Environmental Policy Act (NEPA), to consider potential extension of the Program. On December 1, 2004, a Federal Register Notice was issued approving the program extension. This paper discusses the findings from the NEPA analysis and the potential changes in the program that may result from implementation of the proposed changes. (author)

  1. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

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

    Science.gov (United States)

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

    2007-08-01

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

  3. Effects of the "great recession" on the forest products sector in the northern region of the United States

    Science.gov (United States)

    Christopher W. Woodall; William G. Luppold; Peter J. Ince; Ronald J. Piva; Kenneth E. Skog

    2012-01-01

    The forest industry within the northern region of the United States has demonstrated a notable decline in terms of employment, number of mills, wood consumption, and forest harvests since 2000--a downturn exacerbated by the "Great Recession" of 2007-2009. Longer term industrial decline (since 2000) has been evidenced by reductions in secondary product (e.g.,...

  4. Young Adolescents' Positioning of Human Rights: Findings from Colombia, Northern Ireland, Republic of Ireland and the United States

    Science.gov (United States)

    Barton, Keith C.

    2015-01-01

    This study investigated how young adolescents thought about the location of human rights issues and the nature of violations in differing geographic regions. Open-ended, task-based interviews were conducted with 116 students in Colombia, Northern Ireland, the Republic of Ireland and the United States. Although students in each location pointed to…

  5. 75 FR 44292 - Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2; Notice of...

    Science.gov (United States)

    2010-07-28

    ... and DPR-60] Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2... assessment, and behavioral observation) of the unescorted access authorization program when making the... under consideration to determine whether it met the criteria established in NRC Management Directive (MD...

  6. State formation and water-resource management in the Horn of Africa: the Aksumite Kingdom of the northern Ethiopian Highlands

    DEFF Research Database (Denmark)

    Sulas, Federica; Madella, Marco; French, Charles

    2009-01-01

    Intensification of agriculture and irrigation are often considered triggers for both the flourishing and demise of civilizations. Was irrigation a key factor of state formation and urban development in northern Ethiopia? We argue that a household-based management of farmland and water would have ...... likely that ancient Aksumites utilized seasonal rainfall and water conservation methods to sustain food production....

  7. Use of damage surveys and field inventories to evaluate oak and sugar maple health in the northern United States

    Science.gov (United States)

    Randall S Morin; Christopher W. Woodall; Jim Steinman; Charles H. Perry

    2009-01-01

    Oak species (Quercus spp.) and sugar maple (Acer saccharum) are substantial components of the forest ecosystems in the 24-state region spanning the northern U.S. During recent decades, both damage surveys and forest inventories have documented declines of sugar maple and oak health. In order to more fully assess the status of oak and sugar maple health, we examined...

  8. Responding to the Needs of Young People Leaving State Care: Law, Practice, and Policy in England and Northern Ireland.

    Science.gov (United States)

    Pinkerton, John; Stein, Mike

    1995-01-01

    Notes that the challenge for state child welfare services when young people leave care is to prepare them to cope with pressures surrounding this transition. Reviews existing research to explore current practice in England and Northern Ireland, and considers whether recent legislative reform in the two jurisdictions will help develop policy and…

  9. International topical meeting on research reactor fuel management (RRFM) - United States foreign research reactor (FRR) spent nuclear fuel (SNF) acceptance program: 2010 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.; Niehus, M.T.; Landers, C.

    2010-01-01

    The Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by the United States Department of Energy (DOE), in consultation with the Department of State (DOS) in May 1996, scheduled to expire May 12, 2016, to return research reactor fuel until May 12, 2019 to the U.S. is in its fourteenth year. This paper provides a brief update on the program, part of the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of U.S.-origin nuclear materials, which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible research reactors who have not yet participated in the program. (author)

  10. Modification of the Penn State Reactor to allow transverse and rotational core motion to increase operational versatility

    International Nuclear Information System (INIS)

    Hughes, Daniel E.

    1994-01-01

    At Penn State the Nuclear Engineering students have the opportunity to perform experiments in reactor physics, work with reactor and radiation instrumentation, and operate a nuclear reactor. These activities are done at the Penn State Breazeale Reactor (PSBR), a General Atomics Mark III TRIGA reactor. Unfortunately this activity alone can not fully support the facility. The PSBR is mandated by Penn State to provide a portion of its operating budget by selling services to users outside as well as inside Penn State. In order to increase the marketability of PSBR an upgrade program was started to increase the quality and versatility of operation. The PSBR is the longest operating university reactor in the United States. The first phase of the upgrade program began in 1992. The quality of operation was increased by replacing a 1965 vintage console with a more reliable digital control and monitoring system. The present phase of the upgrade program is to increase the versatility of operation by modifying the reactor to allow transverse and rotational core motion. Adding two more degrees of motion to the reactor core increases the capability of the facility to meet the needs of present and future users. This upgrade is being financed by a grant from the Department of Energy and matching funds from Penn State. (author)

  11. Northern States Power Company's open transmission tariff from a customer's perspective

    International Nuclear Information System (INIS)

    Marietta, K.E.; Achinger, S.K.

    1993-01-01

    In October of 1990, Northern States Power Company (NSP or Company), filed a unique open transmission tariff for both captive customers and through-system transactions. This is an important step towards expanding transmission services in the United States. Many individuals in the utility industry, who may be considering Imposing generation costs on transmission services, have been closely monitoring NSP's case which is currently before the Federal Energy Regulatory Commission (FERC). NSP's innovative generation costs include charges for reactive power production, frequency control, load dispatching, and load following. The results of this case may also have an important impact on the future of open transmission tariffs. Rates for these services depend on the customer's classification as either a captive or through-system consumer. The proposed tariff raises critical issues related to the costing of these transmission services. NSP's methodology has caused serious concern because the proposed tariff would increase transmission costs by an average of 53%. This paper will discuss the benefits of transmission, proposed rates, contract terms, and costing methodologies of NSP's plan

  12. BOVINE NEOSPOROSIS IN CATTLE FARMS FROM THE NORTHERN REGION OF THE STATE OF VERACRUZ, MEXICO

    Directory of Open Access Journals (Sweden)

    Tomás Montiel-Peña

    2011-11-01

    Full Text Available The aim of the study was to determine the presence of antibodies against Neospora caninum and its DNA in blood samples from bovine females from the northern region of the state of Veracruz, Mexico. A cross-sectional epidemiological study was carried out in 13 municipalities, with a sample size of 821 animals. Blood and serum samples were analyzed through ELISA and PCR, respectively. Overall prevalence was 20.8 %; the highest specific prevalences were obtained in breeding cows (27.4 %, crossbred cows (20.9 %, second-calving cows (23.2 %, three year-old cows (20.6 % and cows with abortion history (20 %. The risk factors associated with seropositivity were dairy cattle (OR = 1.9; IC95 %: 1.1-3.4 and dog presence in the farms (OR = 5.3; IC95 %: 1.3-22.3. The presence of N. caninum DNA was demonstrated in 4 out of 12 blood samples tested, which evidenced the existence of active infection. In conclusion, there were risk factors associated with bovine neosporosis, which proved the existence of active infection by N. caninum in cows from the state of Veracruz, Mexico.

  13. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

  14. Current state and prospects of carbon management in high latitudes of Northern Eurasia

    Science.gov (United States)

    Schepaschenko, Dmitry; Shvidenko, Anatoly

    2010-05-01

    The current state and trajectories of future development of natural landscapes in high latitudes of Northern Eurasia are defined inter alia by (1) current unsatisfactory social and economic situation in boreal Northern Eurasia; (2) the dramatic magnitude of on-going and expected climatic change (warming up to 10-12oC under global warming at 4oC); (3) increasing anthropogenic pressure, particularly in regions of intensive oil and gas exploration and extraction; (4) large areas of sparsely populated and practically unmanaged land; (5) vulnerability of northern ecosystems which historically developed under cold climates and buffering capacity of which is not well known; (6) risk of catastrophic natural disturbances (fire, insect outbreaks) whose frequency and severity have accelerated during recent decades; and (7) high probability of irreversible changes of vegetation cover. These specifics are overlapped with insufficient governance of natural renewable resources (e.g., forests) and destructed practice of industrial development of new territories (oil and gas extraction and exploration, metallurgy etc.). Based on a full carbon account for terrestrial vegetation ecosystems of Northern Eurasia, we analyze the relative impacts of major drivers on magnitude and uncertainty of the Net Ecosystem Carbon Balance (NECB) under current and expected climate and environment. Dynamic trends and interannual variability of NECB are mostly dependent on weather conditions during growth seasons of individual years, regimes of natural disturbances, and anthropogenic impacts on ecosystems. In a short term, disturbances and human impacts cause a theoretically 'manageable' part of the full carbon account, which on average is estimated to be of about 20% of annual net primary production. In a long term, thawing of permafrost and change of hydrological regimes of vast territories may result in a catastrophic decline of the forested area and wide distribution of 'green desertification'. The

  15. Primary School Environment Trend, Class-Ratio and Head Teachers Overcrowded Classrooms Management Strategies in Northern Senatorial District of Ondo State, Nigeria

    Science.gov (United States)

    Babatunde, Ehinola Gabriel

    2015-01-01

    Primary school Enrolment Trend, Class-Ratio and Head Teachers overcrowded classrooms management strategies in Northern Senatorial District of Ondo State, Nigeria was investigated. The purpose of the study is to examine the current enrolment trend in public primary schools in northern senatorial District of Ondo State. Also, is to ascertain the…

  16. New record of the mangrove rivulid Kryptolebias hermaphroditus Costa, 2011 (Cyprinodontiformes: Cynolebiidae) in the Pará state, northern Brazil

    OpenAIRE

    Guimarães-Costa, Auryceia; Schneider, Horacio; Sampaio, Iracilda

    2017-01-01

    The mangrove killifish Kryptolebias herma­phro­ditus is reported to the southeast and northeast regions of Brazil. Recently, a specimen of K. hermaphroditus was collected in a shallow running seawater stream at Ajuruteua beach, Pará state, northern Brazil. This new record is ca. 1,350 km from the nearest previously known occurrence in Rio Grande do Norte state, Brazil. Morphological, molecular, and ecological aspects of this species are described.

  17. Status of oak seedlings and saplings in the northern United States: implications for sustainability of oak forests

    Science.gov (United States)

    Chris W. Woodall; Randall S. Morin; Jim R. Steinman; Charles H. Perry

    2008-01-01

    Oak species are a substantial component of forest ecosystems in a 24-state region spanning the northern U.S. During recent decades, it has been documented that the health of oak forests has been experiencing large-scale decline. To further evaluate the sustainability of oak forests in nearly half the states of the U.S., the current status of oak seedlings and saplings...

  18. Applications of Oregon State University's TRIGA reactor in health physics education

    International Nuclear Information System (INIS)

    Higginbotham, J.F.

    1990-01-01

    The Oregon State University TRIGA reactor (OSTR) is used to support a broad range of traditional academic disciplines, including anthropology, oceanography, geology, physics, nuclear chemistry, and nuclear engineering. However, it also finds extensive application in the somewhat more unique area of health physics education and research. This paper summarizes these health physics applications and briefly describes how the OSTR makes important educational contributions to the field of health physics

  19. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1989-01-01

    The United States has made substantial progress in achieving LMR programme objectives. A decision was made in 1988 to select the General Electric ALMR concept known as PRISM (Power Reactor Innovative Safe Module) for advanced conceptual design. A 3-year contract was awarded to General Electric in January of this year for concentrated trade-off studies and advanced design development. The strategy is to integrate those advancements that best meet programme objectives into a national ALMR system concept. (author). 8 figs

  20. Emerging Capripoxvirus disease outbreaks in Himachal Pradesh, a northern state of India.

    Science.gov (United States)

    Verma, S; Verma, L K; Gupta, V K; Katoch, V C; Dogra, V; Pal, B; Sharma, M

    2011-02-01

    Both sheep and goat pox are contagious viral diseases and affect small ruminants and are caused by sheep pox virus and goat pox virus respectively that belong to genus Capripoxvirus of Poxviridae family. Huge economic losses emanating from the disease outbreaks are the results of the wool and hide damage, subsequent production losses and also the morbidities and mortalities associated with the disease. This communication highlights clinico-epidemiological observations from the two sheep pox and one goat pox outbreaks. Grossly, multisystemic nodular lesions, mucopurulent nasal discharges and respiratory symptoms were observed in the affected animals. The morbidity, mortality and case fatality rates were 5.18%, 2.45% and 32.37%, respectively. Histopathological, haematological, molecular and serological techniques and also isolation of virus in embryonated chicken eggs were used for the diagnosis of the diseases. The spatial distribution of the disease signifies the role of common pasturelands used for grazing the animals while temporally all three outbreaks occurred in winters and were probably associated with cold stress and fodder scarcity. This is the first recorded report of Capripoxvirus infection in recent times and it highlights the disease as one of the emerging diseases in the northern state of Himachal Pradesh in India. © 2010 Blackwell Verlag GmbH.

  1. A Best-Estimate Reactor Core Monitor Using State Feedback Strategies to Reduce Uncertainties

    International Nuclear Information System (INIS)

    Martin, Robert P.; Edwards, Robert M.

    2000-01-01

    The development and demonstration of a new algorithm to reduce modeling and state-estimation uncertainty in best-estimate simulation codes has been investigated. Demonstration is given by way of a prototype reactor core monitor. The architecture of this monitor integrates a control-theory-based, distributed-parameter estimation technique into a production-grade best-estimate simulation code. The Kalman Filter-Sequential Least-Squares (KFSLS) parameter estimation algorithm has been extended for application into the computational environment of the best-estimate simulation code RELAP5-3D. In control system terminology, this configuration can be thought of as a 'best-estimate' observer. The application to a distributed-parameter reactor system involves a unique modal model that approximates physical components, such as the reactor, by describing both states and parameters by an orthogonal expansion. The basic KFSLS parameter estimation is used to dynamically refine a spatially varying (distributed) parameter. The application of the distributed-parameter estimator is expected to complement a traditional nonlinear best-estimate simulation code by providing a mechanism for reducing both code input (modeling) and output (state-estimation) uncertainty in complex, distributed-parameter systems

  2. Monticello Mill Tailings Site Operable Unit III Annual Groundwater Report May 2014 Through April 2015, October 2015

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Jason [USDOE Office of Legacy Management, Washington, DC (United States); Smith, Fred [Navarro Research and Engineering, Inc., Oak Ridge, TN (United States)

    2015-10-01

    This report provides the annual analysis of water quality restoration progress, cumulative through April 2015, for Operable Unit (OU) III, surface water and groundwater, of the U.S. Department of Energy (DOE) Office of Legacy Management Monticello Mill Tailings Site (MMTS). The MMTS is a Comprehensive Environmental Response, Compensation, and Liability Act National Priorities List site located in and near the city of Monticello, San Juan County, Utah. MMTS comprises the 110-acre site of a former uranium- and vanadium-ore-processing mill (mill site) and 1,700 acres of surrounding private and municipal property. Milling operations generated 2.5 million cubic yards of waste (tailings) from 1942 to 1960. The tailings were impounded at four locations on the mill site. Inorganic constituents in the tailings drained from the impoundments to contaminate local surface water (Montezuma Creek) and groundwater in the underlying alluvial aquifer. Mill tailings dispersed by wind and water also contaminated properties surrounding and downstream of the mill site. Remedial actions to remove and isolate radiologically contaminated soil, sediment, and debris from the former mill site (OU I) and surrounding properties (OU II) were completed in 1999 with the encapsulation of the wastes in an engineered repository located on DOE property 1 mile south of the former mill site. Contamination of groundwater and surface water remains within OU III at levels that exceed water quality protection standards. Uranium is the primary contaminant of concern. LM implemented monitored natural attenuation with institutional controls as the OU III remedy in 2004. Because groundwater restoration proceeded more slowly than expected and did not meet performance criteria established in the OU III Record of Decision (June 2004), LM implemented a contingency action in 2009 by an Explanation of Significant Difference to include a pump-and-treat system using a single extraction well and treatment by zero

  3. University research reactors in the United States: Their role and value

    International Nuclear Information System (INIS)

    1988-01-01

    This report is primarily addressed to the people who make decisions affecting the levels of future university reactor programs URR: university administrators, department heads, federal policy makers, state and local policy makers, those in industry and government who depend upon a supply of nuclear-trained personnel, and those who are concerned with the future of the many sciences that benefit from the unique capabilities of nuclear-based techniques as well as from the nuclear sciences themselves. The major thrust of this report is to illustrate the scientific and social benefits and contributions associated with well-managed and well-funded university reactor programs. The intent is to help a decision maker gain a perspective and appreciation of the scientific, academic, social, and technical values of URR programs. The report also examines the role of university-like reactors in Europe, where a productive community of researchers is apparently served in an exemplary manner. The committee, assesses the security and safeguard needs at small reactors in a university setting in order to help gain a perspective on the potential hazards and relative risks involved. The last chapter discusses the kind of commitment and support needed if a significant population of URRs is to remain productive. 83 refs., 11 figs., 2 tabs

  4. A Neural-Network-Based Nonlinear Adaptive State-Observer for Pressurized Water Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2013-10-01

    Full Text Available Although there have been some severe nuclear accidents such as Three Mile Island (USA, Chernobyl (Ukraine and Fukushima (Japan, nuclear fission energy is still a source of clean energy that can substitute for fossil fuels in a centralized way and in a great amount with commercial availability and economic competitiveness. Since the pressurized water reactor (PWR is the most widely used nuclear fission reactor, its safe, stable and efficient operation is meaningful to the current rebirth of the nuclear fission energy industry. Power-level regulation is an important technique which can deeply affect the operation stability and efficiency of PWRs. Compared with the classical power-level controllers, the advanced power-level regulators could strengthen both the closed-loop stability and control performance by feeding back the internal state-variables. However, not all of the internal state variables of a PWR can be obtained directly by measurements. To implement advanced PWR power-level control law, it is necessary to develop a state-observer to reconstruct the unmeasurable state-variables. Since a PWR is naturally a complex nonlinear system with parameters varying with power-level, fuel burnup, xenon isotope production, control rod worth and etc., it is meaningful to design a nonlinear observer for the PWR with adaptability to system uncertainties. Due to this and the strong learning capability of the multi-layer perceptron (MLP neural network, an MLP-based nonlinear adaptive observer is given for PWRs. Based upon Lyapunov stability theory, it is proved theoretically that this newly-built observer can provide bounded and convergent state-observation. This observer is then applied to the state-observation of a special PWR, i.e., the nuclear heating reactor (NHR, and numerical simulation results not only verify its feasibility but also give the relationship between the observation performance and observer parameters.

  5. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1982-August 1983

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1984-03-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1982 was over 5500 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1982 through August 1983

  6. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1981-August 1982

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1982-12-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1981 and nearly 5000 megawatt-hours) research reactor in the mid-Atlantic States. In addition, a second, small (50 watt) reactor is also available for use in educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1981 through August 1982

  7. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1983-August 1984

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.

    1984-11-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1983 was over 6000 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1983 through August 1984

  8. Status of liquid metal reactor development in the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Griffith, Jerry D [Reactor Systems Development and Technology, Office of Nuclear Energy, U.S. Department of Energy (United States); Horton, Kenneth E [Division of International Programs, Office of Nuclear Energy, U.S. Department of Energy (United States)

    1992-07-01

    The U.S. civilian nuclear power research and development program continues to focus on advanced large and mid-size light water reactors, advanced liquid metal fast reactors, and modular high temperature gas cooled reactors. This paper discusses the Advanced Liquid Metal Reactor program, which is composed of a small, passively safe fast reactor coupled with a metal fuel cycle that incorporates actinide recycle, and an emerging effort to process LWR spent fuel for LMR fissile material, and to enhance the LWR waste management. The liquid metal reactor concept has a sound technology base, with some three decades of research and development both in this and other countries. An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Notable among the research facilities is the Experimental Breeder Reactor-Il (EBR-II) at Argonne National Laboratory (ANL) in Idaho and the Fast Flux Test Facility (FFTF) at Hanford, Washington. Subsequent to the DOE directive to shut down the Fast Flux Test Facility in early 1990, significant effort was placed in finding international financial support for this reactor. This initiative was not successful. Therefore, although there may be a potential future mission for the FFTF, the Secretary of Energy announced on March 13, 1992 that the FFTF will be put in a standby condition starting April 1, 1992. Current U.S. Advanced Liquid Metal Reactor (ALMR) activity is focused on providing a reactor and fuel cycle system with improved safety margins, better economics, and an attractive waste management (actinide recycle) option. Special attention is being directed to passive safety features, large design margins, modular plant construction, standardized plant design leading to simplified licensing and shorter construction schedules, factory fabrication

  9. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, Jerry D.; Horton, Kenneth E.

    1992-01-01

    The U.S. civilian nuclear power research and development program continues to focus on advanced large and mid-size light water reactors, advanced liquid metal fast reactors, and modular high temperature gas cooled reactors. This paper discusses the Advanced Liquid Metal Reactor program, which is composed of a small, passively safe fast reactor coupled with a metal fuel cycle that incorporates actinide recycle, and an emerging effort to process LWR spent fuel for LMR fissile material, and to enhance the LWR waste management. The liquid metal reactor concept has a sound technology base, with some three decades of research and development both in this and other countries. An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Notable among the research facilities is the Experimental Breeder Reactor-Il (EBR-II) at Argonne National Laboratory (ANL) in Idaho and the Fast Flux Test Facility (FFTF) at Hanford, Washington. Subsequent to the DOE directive to shut down the Fast Flux Test Facility in early 1990, significant effort was placed in finding international financial support for this reactor. This initiative was not successful. Therefore, although there may be a potential future mission for the FFTF, the Secretary of Energy announced on March 13, 1992 that the FFTF will be put in a standby condition starting April 1, 1992. Current U.S. Advanced Liquid Metal Reactor (ALMR) activity is focused on providing a reactor and fuel cycle system with improved safety margins, better economics, and an attractive waste management (actinide recycle) option. Special attention is being directed to passive safety features, large design margins, modular plant construction, standardized plant design leading to simplified licensing and shorter construction schedules, factory fabrication

  10. Progress of design studies on an LHD-type steady-state reactor

    International Nuclear Information System (INIS)

    Motojima, O.; Komori, A.; Sagara, A.

    2007-01-01

    Helical Heliotrons such as the Large Helical Device (LHD) and Stellarators (H and S systems) have a high potential to realize a current-less steady-state and stable magnetic fusion energy reactor as an alternative to the tokamak DEMO-reactor. H and S systems ideally have an intrinsic property of Q=infinite. Here it is very important to remember that the understanding of the physics of 3-D toroidal magnetic confinement system is naturally extended to tokamak systems. The physics is universal among these two types of systems and the technology is common. We present our recent results from LHD experiments and reactor studies of a next generation LHD-type DEMO Reactor called FFHR. (1) Development of 3-D superconducting (SC) coil technology Due to the successful results of the LHD construction from 1990 to 2007, and steady operation over 8 years from 1998 to 2007, more than 2,000 hrs/year at a high field of around 3 Tesla, we have a large enough data base to demonstrate that 3D coil technology has become the standard technology for a fusion energy reactor. LHD is the largest SC fusion device in the world, contributing to the development of the SC technology necessary for fusion research. The poloidal coils of LHD adopted a super critical forced flow cooling system and their dimensions are almost the same as the ITER toroidal coils. (2) Extended physics understanding of high beta, high T, high n τT , and steady state operation Recent LHD experiments have demonstrated the broad and advanced capabilities of LHD as a toroidal magnetic confinement device, which are highlighted by the achievements of 5% volume averaged beta, electron and ion temperatures of 10 keV, super high density of 10E15/cc and 1 hr discharges. We plan to increase the heating power up to 35 MW, and to use deuterium gas for confinement improvement. The n τT will be improved to the design nominal value of Q=0.3 within several years and ultimately would approach unity. The key issue for this is the

  11. Manager's handbook for northern hardwoods in the north-central states.

    Science.gov (United States)

    Carl H. Tubbs

    1977-01-01

    Provides a key for the resource manager to use in choosing silvicultural practices for the management of northern hardwoods. Control of stand composition, growth, and stand establishment for timber production, water, wildlife, and recreation are discussed.

  12. Classification of Reactor Facility Operational State Using SPRT Methods with Radiation Sensor Networks

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez Aviles, Camila A. [ORNL; Rao, Nageswara S. [ORNL

    2018-01-01

    We consider the problem of inferring the operational state of a reactor facility by using measurements from a radiation sensor network, which is deployed around the facility’s ventilation stack. The radiation emissions from the stack decay with distance, and the corresponding measurements are inherently random with parameters determined by radiation intensity levels at the sensor locations. We fuse measurements from network sensors to estimate the intensity at the stack, and use this estimate in a one-sided Sequential Probability Ratio Test (SPRT) to infer the on/off state of the reactor facility. We demonstrate the superior performance of this method over conventional majority vote fusers and individual sensors using (i) test measurements from a network of NaI sensors, and (ii) emulated measurements using radioactive effluents collected at a reactor facility stack. We analytically quantify the performance improvements of individual sensors and their networks with adaptive thresholds over those with fixed ones, by using the packing number of the radiation intensity space.

  13. The Intense Slow Positron Beam Facility at the NC State University PULSTAR Reactor

    International Nuclear Information System (INIS)

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

    2009-01-01

    An intense slow positron beam is in its early stages of operation at the 1-MW open-pool PULSTAR research reactor at North Carolina State University. The positron beam line is installed in a beam port that has a 30-cmx30-cm cross sectional view of the core. The positrons are created in a tungsten converter/moderator by pair-production using gamma rays produced in the reactor core and by neutron capture reactions in cadmium cladding surrounding the tungsten. Upon moderation, slow (∼3 eV) positrons that are emitted from the moderator are electrostatically extracted, focused and magnetically guided until they exit the reactor biological shield with 1-keV energy, approximately 3-cm beam diameter and an intensity exceeding 6x10 8 positrons per second. A magnetic beam switch and transport system has been installed and tested that directs the beam into one of two spectrometers. The spectrometers are designed to implement state-of-the-art PALS and DBS techniques to perform positron and positronium annihilation studies of nanophases in matter.

  14. Steady-state resistive toroidal-field coils for tokamak reactors

    International Nuclear Information System (INIS)

    Kalnavarns, J.; Jassby, D.L.

    1979-12-01

    If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems

  15. Wave-driver options for low-aspect-ratio steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1981-02-01

    Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value

  16. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures.

    Science.gov (United States)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R; Crowhurst, Jonathan C; Weisz, David G; Zaug, Joseph M; Dai, Zurong; Radousky, Harry B; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L; Cappelli, Mark A; Rose, Timothy P

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  17. Ten-year utilization of the Oregon State University TRIGA Reactor (OSTR)

    International Nuclear Information System (INIS)

    Ringle, John C.; Anderson, Terrance V.; Johnson, Arthur G.

    1978-01-01

    The Oregon State University TRIGA Reactor (OSTR) has been used heavily throughout the past ten years to accommodate exclusively university research, teaching, and training efforts. Averages for the past nine years show that the OSTR use time has been as follows: 14% for academic and special training courses; 44% for OSU research projects; 6% for non-OSU research projects; 2% for demonstrations for tours; and 34% for reactor maintenance, calibrations, inspections, etc. The OSTR has operated an average of 25.4 hours per week during this nine-year period. Each year, about 20 academic courses and 30 different research projects use the OSTR. Visitors to the facility average about 1,500 per year. No commercial radiations or services have been performed at the OSTR during this period. Special operator training courses are given at the OSTR at the rate of at least one per year. (author)

  18. Chagas disease ecoepidemiology and environmental changes in northern Minas Gerais state, Brazil.

    Science.gov (United States)

    Vianna, Elisa Neves; Souza E Guimarães, Ricardo José de Paula; Souza, Christian Rezende; Gorla, David; Diotaiuti, Liléia

    2017-11-01

    Triatoma sordida and Triatoma pseudomaculata are frequently captured triatomine species in the Brazilian savannah and caatinga biomes, respectively, and in Brazilian domiciles. This study identified eco-epidemiological changes in Chagas disease in northern Minas Gerais state, Brazil, and considered the influence of environmental shifts and both natural and anthropogenic effects. Domicile infestation and Trypanosoma cruzi infection rates were obtained from triatomines and sylvatic reservoirs during the following two time periods: the 1980s and 2007/2008. Entomological and climatic data with land cover classification derived from satellite imagery were integrated into a geographic information system (GIS), which was applied for atmospheric correction, segmentation, image classification, and mapping and to analyse data obtained in the field. Climatic data were analysed and compared to land cover classifications. A comparison of current data with data obtained in the 1980's showed that T. sordida colonised domiciliary areas in both periods, and that T. pseudomaculata did not colonise these areas. There was a tendency toward a reduction in T. cruzi infection rates in sylvatic reservoirs, and of triatomines captured in both households and in the sylvatic environment. T. sordida populations have reduced in the sylvatic environment, while T. pseudomaculata showed an expanding trend in the region compared to counts observed in the 1980's in the sylvatic environment. This may be related to high deforestation rates as well as gradual increases in land surface temperature (LST) and temperatures along the years. Our results suggest a geographical expansion of species into new biomes as a result of anthropogenic and climatic changes that directly interfere with the reproductive and infection processes of vectors.

  19. Chagas disease ecoepidemiology and environmental changes in northern Minas Gerais state, Brazil

    Directory of Open Access Journals (Sweden)

    Elisa Neves Vianna

    Full Text Available BACKGROUND Triatoma sordida and Triatoma pseudomaculata are frequently captured triatomine species in the Brazilian savannah and caatinga biomes, respectively, and in Brazilian domiciles. OBJECTIVES This study identified eco-epidemiological changes in Chagas disease in northern Minas Gerais state, Brazil, and considered the influence of environmental shifts and both natural and anthropogenic effects. METHODS Domicile infestation and Trypanosoma cruzi infection rates were obtained from triatomines and sylvatic reservoirs during the following two time periods: the 1980s and 2007/2008. Entomological and climatic data with land cover classification derived from satellite imagery were integrated into a geographic information system (GIS, which was applied for atmospheric correction, segmentation, image classification, and mapping and to analyse data obtained in the field. Climatic data were analysed and compared to land cover classifications. RESULTS A comparison of current data with data obtained in the 1980's showed that T. sordida colonised domiciliary areas in both periods, and that T. pseudomaculata did not colonise these areas. There was a tendency toward a reduction in T. cruzi infection rates in sylvatic reservoirs, and of triatomines captured in both households and in the sylvatic environment. T. sordida populations have reduced in the sylvatic environment, while T. pseudomaculata showed an expanding trend in the region compared to counts observed in the 1980's in the sylvatic environment. This may be related to high deforestation rates as well as gradual increases in land surface temperature (LST and temperatures along the years. MAIN CONCLUSIONS Our results suggest a geographical expansion of species into new biomes as a result of anthropogenic and climatic changes that directly interfere with the reproductive and infection processes of vectors.

  20. Different Stratospheric Polar Vortex States linked to Cold-Spells in North America and Northern Eurasia

    Science.gov (United States)

    Kretschmer, M.; Cohen, J. L.; Runge, J.; Coumou, D.

    2017-12-01

    The stratospheric polar vortex in boreal winter can influence the tropospheric circulation and thereby surface weather in the mid-latitudes. Weak states of the vortex, e.g. associated with Sudden Stratospheric Warmings (SSWs), often precede a negative phase of the North Atlantic Oscillation (NAO), and thus increase the risk of mid-latitude cold-spells especially over Eurasia. Here we show using cluster analysis that next to the well-documented relationship between a zonally symmetric disturbed vortex and a negative NAO, there exists a zonally asymmetric pattern linked to a negative Western Pacific Oscillation (WPO) and cold-spells in the northeastern US, like for example observed in February 2014. The latter is more synoptic in time-scale but occurs more frequently than SSWs. A causal effect network (CEN) approach gives insights into the underlying physical pathways and time-lags showing that high-pressure around Greenland leads to vertical wave activity over eastern Siberia leading to downward propagating waves over Alaska and high pressure over the North Pacific. Moreover, composites propose that a rather strong mid-stratospheric vortex seems to be favorable for this zonally asymmetric and reflective mechanism. Overall, the mutual relationship between stratospheric circulation and high-latitude blocking in both the Pacific and Atlantic Oceans is complex and involves mechanisms operating at different time-scales. Our results suggest that the stratospheric influence on winter circulation should not exclusively be analyzed in terms of a downward propagating Northern Annular Mode (NAM) signal and SSWs. In particular when studying the stratospheric impacts on North American temperature it is crucial to also consider the more transient and zonally asymmetric events which might help to improve seasonal winter predictions for this region.

  1. The ethnoecological knowledge of fishermen from three coastal lagoons in the northern of the State of Rio de Janeiro, Brazil

    OpenAIRE

    Lopes, Alexandre Ferreira; Bozelli, Reinaldo Luiz

    2014-01-01

    The current study investigated the ethnoecological knowledge developed by fishermen through their fishing activities and searched for ways to match such knowledge to empirical data available in the scientific literature. The research involved fishermen from three coastal lagoons in the northern region of the State of Rio de Janeiro, Brazil, who were consulted through semi-structured interviews after the establishment of a trustful relationship over a period of three years with the interviewer...

  2. Detection of the pandemic norovirus variant GII.4 Sydney 2012 in Rio Branco, state of Acre, northern Brazil

    Directory of Open Access Journals (Sweden)

    Luciana Damascena da Silva

    2013-12-01

    Full Text Available Noroviruses (NoVs are important cause of gastroenteritis in humans worldwide. Genotype GII.4 is responsible for the majority of outbreaks reported to date. This study describes, for the first time in Brazil, the circulation of NoV GII.4 variant Sydney 2012 in faecal samples collected from children aged less than or equal to eight years in Rio Branco, state of Acre, northern Brazil, during July-September 2012.

  3. Safety Evaluation Report related to the renewal of the operating license for the research reactor at Pennsylvania State University

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Evaluation Report for the application filed by the Pennsylvania State University for a renewal of Operating License R-2 to continue to operate the Pennsylvania State University Breazeale Reactor (PSBR) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the campus in University Park, Pennsylvania. On the basis of its technical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the environment

  4. Politika «Severnogo izmerenija»: sovremennoe sostojanie i perspektivy razvitija [The Northern Dimension policy: current state and development prospects

    Directory of Open Access Journals (Sweden)

    Bolotnikova Yekaterina

    2010-01-01

    Full Text Available This article examines the evolution and current state of the Northern Dimension policy and its role in Russia-EU relations. The authors analyse the discrepancy between the actual achievements of the Northern Dimension and its potential and the over-high expectations, which accompanied the policy renewal.

  5. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave rf energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected rf energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected rf energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range delta . The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width delta in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma

  6. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    International Nuclear Information System (INIS)

    Bers, A.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave rf energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected rf energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected rf energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range delta . The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width delta in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma

  7. Development of an aging evaluation and life extension program for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Dwight, J.E. Jr.

    1988-01-01

    A life extension program has been developed for the US Department of Energy's Advanced Test Reactor. The program is an adaptation of life extension pilot programs at the Surry Unit 1 and Monticello generating stations and is being completed in three phases. In Phase 1, the critical plant components were identified. In Phase 2, existing lifetime analyses and support data for the critical components were reviewed. The results from the review give a preliminary indication that an overall plant lifetime in excess of forty years is feasible. In Phase 3, now in progress, detailed evaluations for component life extensions are being performed. 2 refs., 2 figs., 1 tab

  8. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  9. Radiotherapy in Northern Germany: Facts and figures about radiooncology in 6 states

    International Nuclear Information System (INIS)

    Brodersen, H.J.; Heilmann, H.P.

    1996-01-01

    Traditionally there was a close cooperation between the Roentgen Society of Northern Germany and the Society of 'Niedersachsen' and 'Sachsen-Anhalt'. Therefore, hospital departments and free standing office-type facilities in radiooncology from 6 states have put together a regional North German data-base. Detailed questionnaires were sent out in order to get data on heads of departments, professional staff, technical equipment, workload, diagnoses and training facilities. The collected data were distributed to all participating centers for confidential individual analysis. This paper reports the data collected. The survey 1995/96 had an unexpected return. Thirty-six from 38 centers in a region with 17.36 million inhabitants in 6 states (21.3% of Germany) took part. There were 73 megavolt machines and 31 afterloading facilities. Two hundred and forty physicians, 110 physicists, and 350 radiographers were working in radiooncology; and there were 852 hospital beds specifically assigned to radiooncology for the care of in-patients. In 1995, 32,000 patients were treated with 37,000 series, 570,000 treatment visits and 1,600,000 radiation fields. The average patient got 1.2 series, 18 treatment visits and 50 fields. Equipmentwise, an average department runs a mean of 2 megavolt machines and 1 afterloading facility. The corresponding data for 1 megavolt machine were 3.5 physicians, 1.6 physicists, 5 radiographers and 13 hospital beds. There was 1 machine for 240,000 inhabitants. The average workload of a megavolt machine was 439 patients a year with 513 series, 7,813 treatment visits and 21,845 radiation fields. A regional data-base for radiooncology is being described. It is demonstrated that it is possible to get detailed and current data by this method. It provides transparent data on equipment, personnel and workload in radiooncology. It should set an example for other regions to collect their data in the same fashion, and then compile all such data for the whole

  10. A highly efficient autothermal microchannel reactor for ammonia decomposition: Analysis of hydrogen production in transient and steady-state regimes

    Science.gov (United States)

    Engelbrecht, Nicolaas; Chiuta, Steven; Bessarabov, Dmitri G.

    2018-05-01

    The experimental evaluation of an autothermal microchannel reactor for H2 production from NH3 decomposition is described. The reactor design incorporates an autothermal approach, with added NH3 oxidation, for coupled heat supply to the endothermic decomposition reaction. An alternating catalytic plate arrangement is used to accomplish this thermal coupling in a cocurrent flow strategy. Detailed analysis of the transient operating regime associated with reactor start-up and steady-state results is presented. The effects of operating parameters on reactor performance are investigated, specifically, the NH3 decomposition flow rate, NH3 oxidation flow rate, and fuel-oxygen equivalence ratio. Overall, the reactor exhibits rapid response time during start-up; within 60 min, H2 production is approximately 95% of steady-state values. The recommended operating point for steady-state H2 production corresponds to an NH3 decomposition flow rate of 6 NL min-1, NH3 oxidation flow rate of 4 NL min-1, and fuel-oxygen equivalence ratio of 1.4. Under these flows, NH3 conversion of 99.8% and H2 equivalent fuel cell power output of 0.71 kWe is achieved. The reactor shows good heat utilization with a thermal efficiency of 75.9%. An efficient autothermal reactor design is therefore demonstrated, which may be upscaled to a multi-kW H2 production system for commercial implementation.

  11. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  12. Present state of inspection robot technology in nuclear power facilities. Case of fast breeder reactors

    International Nuclear Information System (INIS)

    Ara, Kuniaki

    1995-01-01

    In the maintenance works in nuclear power facilities such as checkup, inspection and repair, for the main purpose of radiation protection, remote operation technology was introduced since relatively early stage, and at present, the robots that carry out the inspection works for confirming the soundness of main equipment have been developed and put to practical use. At the time of introducing these technologies, in addition to the research and development of robots proper, the coordination with the design of plant machinery and equipment facilities as the premise of introducing robots is an important requirement. In this report, the present state of the development of remote inspection technology for fast breeder reactors is introduced, and the matters to which attention is paid in the plant design for introducing robots are explained. First, fast breeder reactors are described. The needs of robotizing and adopting remote operation in nuclear power facilities are explained, using the examples of the inspection system for a reactor vessel and the inspection system for steam generator heat transfer tubes. (K.I.)

  13. Look at potential options for the fast reactor fuel cycle in the United States

    International Nuclear Information System (INIS)

    Burch, W.D.

    1984-01-01

    This paper reviews the status and plans for the fast reactor fuel cycle in the United States, presents some options that are under consideration, and describes how these options are being evaluated at the present time. The United States will undertake some far-reaching examinations of the entire breeder program strategy in the coming year, and the outcome of these reviews cannot be predicted today. In other papers at this conference you have heard various perspectives from both government and industry representatives. The proposed studies to examine the associated fuel cycle strategies as they relate to the overall emerging breeder strategy are described. The present status of and recent developments in the fuel cycle R and D programs will also be summarized and updated in order to present an overall picture of the United States situation

  14. Steady state characteristics of acclimated hydrogenotrophic methanogens on inorganic substrate in continuous chemostat reactors.

    Science.gov (United States)

    Ako, Olga Y; Kitamura, Y; Intabon, K; Satake, T

    2008-09-01

    A Monod model has been used to describe the steady state characteristics of the acclimated mesophilic hydrogenotrophic methanogens in experimental chemostat reactors. The bacteria were fed with mineral salts and specific trace metals and a H(2)/CO(2) supply was used as a single limited substrate. Under steady state conditions, the growth yield (Y(CH4)) reached 11.66 g cells per mmol of H(2)/CO(2) consumed. The daily cells generation average was 5.67 x 10(11), 5.25 x 10(11), 4.2 x 10(11) and 2.1 x 10(11) cells/l-culture for the dilutions 0.071/d, 0.083/d, 0.1/d and 0.125/d, respectively. The maximum specific growth rate (mu(max)) and the Monod half-saturation coefficient (K(S)) were 0.15/d and 0.82 g/L, respectively. Using these results, the reactor performance was simulated. During the steady state, the simulation predicts the dependence of the H(2)/CO(2) concentration (S) and the cell concentration (X) on the dilution rate. The model fitted the experimental data well and was able to yield a maximum methanogenic activity of 0.24 L CH(4)/g VSS.d. The dilution rate was estimated to be 0.1/d. At the dilution rate of 0.14/d, the exponential cells washout was achieved.

  15. Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

    International Nuclear Information System (INIS)

    Olynyk, G.M.; Hartwig, Z.S.; Whyte, D.G.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Mumgaard, R.T.; Podpaly, Y.A.

    2012-01-01

    Highlights: ► A new scaling for obtaining reactor similarity in the divertor of scaled tokamaks. ► Conceptual design for a tokamak (“Vulcan”) to implement this new scaling. ► Demountable superconducting coils and compact neutron shielding. ► Helium-cooled high-temperature vacuum vessel and first wall. ► High-field-side lower hybrid current drive for non-inductive operation. - Abstract: An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ≃ 1 MW m −2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ∼ R −2/7 . While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system

  16. Panel session on the state of the art in nuclear reactor technology

    International Nuclear Information System (INIS)

    Roche, R.

    1977-01-01

    The state of the art in the technology of pressure vessels and piping of the primary cooling circuit of nuclear steam supply systems is discussed. Design and analysis are considered in the frame of the two types of nuclear reactor retained in France (PWR and the pool type LMFBR). Designing nuclear pressure vessels asks for some more specific Codes and Standards than for conventional vessels, and the stress analysis complementing by a direct comparison between operating loads and failure loads is a mandatory practice in France. As for pool type LMFBR, the structural problems of the nuclear vessel are essentially due to component shape, small thickness, and large stress range

  17. Current state of the auto-evaluation process of the behaviour code in the safety of research reactors in Mexico

    International Nuclear Information System (INIS)

    Mamani A, Y. R.; Salgado G, J. R.

    2011-11-01

    In Mexico, the regulator organism in nuclear matter is the National Commission of Nuclear Safety and Safeguards, and a nuclear research reactor exists, the TRIGA Mark III, operated by the National Institute of Nuclear Research. In this work the main aspects of the current state and the future challenges are presented with relationship to the installation of the auto-evaluation process of the behaviour code in the safety of research reactors for the TRIGA reactor case. Additionally, the legal mark of the licensing process for the nuclear activities in a research reactor is described in a brief way, and the main characteristics of the reactor, the uses for the isotopes production, the administration and the verification of the safety, the administration program of the radiological protection, the emergency plan and the operation personnel qualification are emphasized. (Author)

  18. Mathematical Modeling and Simulation of the Dehydrogenation of Ethyl Benzene to Form Styrene Using Steady-State Fixed Bed Reactor

    Directory of Open Access Journals (Sweden)

    Zaidon M. Shakoor

    2013-05-01

    Full Text Available In this research, two models are developed to simulate the steady state fixed bed reactor used for styrene production by ethylbenzene dehydrogenation. The first is one-dimensional model, considered axial gradient only while the second is two-dimensional model considered axial and radial gradients for same variables.The developed mathematical models consisted of nonlinear simultaneous equations in multiple dependent variables. A complete description of the reactor bed involves partial, ordinary differential and algebraic equations (PDEs, ODEs and AEs describing the temperatures, concentrations and pressure drop across the reactor was given. The model equations are solved by finite differences method. The reactor models were coded with Mat lab 6.5 program and various numerical techniques were used to obtain the desired solution.The simulation data for both models were validated with industrial reactor results with a very good concordance.

  19. Bioindicators of climate and trophic state in lowland and highland aquatic ecosystems of the Northern Neotropics

    Directory of Open Access Journals (Sweden)

    Liseth Pérez

    2013-06-01

    Full Text Available Chironomids, diatoms and microcrustaceans that inhabit aquatic ecosystems of the Northern Neotropics are abundant and diverse. Some species are highly sensitive to changes in water chemical composition and trophic state. This study was undertaken as a first step in developing transfer functions to infer past environmental conditions in the Northern lowland Neotropics. Bioindicator species abundances were related to multiple environmental variables to exploit their use as environmental and paleoenvironmental indicators. We collected and analyzed water and surface sediment samples from 63 waterbodies located along a broad trophic state gradient and steep gradients of altitude (~0-1 560m.a.s.l. and precipitation (~400-3 200mm/y, from NW Yucatán Peninsula (Mexico to southern Guatemala. We related 14 limnological variables to relative abundances of 282 diatom species, 66 chironomid morphospecies, 51 species of cladocerans, 29 non-marine ostracode species and six freshwater calanoid copepods. Multivariate statistics indicated that bicarbonate is the strongest driver of chironomid and copepod distribution. Trophic state is the second most important factor that determines chironomid distribution. Conductivity, which is related to the precipitation gradient and marine influence on the Yucatán Peninsula, is the main variable that shapes diatom, ostracode and cladoceran communities. Diatoms, chironomids and cladocerans displayed higher diversities (H=2.4-2.6 than ostracodes and copepods (H=0.7- 1.8. Species richness and diversity were greater at lower elevations (Los quironómidos, diatomeas y microcrustaceos que habitan ecosistemas acuáticos en el norte de los Neotrópicos son abundantes y diversos. Algunas especies son altamente sensibles a cambios en la composición química del agua y en el estado trófico. Este estudio se realizó como el primer paso para desarrollar funciones de transferencia para inferir condiciones ambientales en el norte de las

  20. United States Army Counter Partisan Operations in Northern Virginia During the American Civil War

    Science.gov (United States)

    2016-06-10

    retribution against civilians occurred on August 21, 1863, when following an ambush on a Union picket site, Brevet Brigadier General George A. Custer...Additionally, northern newspapers reported that these “vandal acts” of retribution were not effective in deterring the partisans, nor gaining

  1. INVENTORY AND DESCRIPTION OF COMMERCIAL REACTOR FUELS WITHIN THE UNITED STATES

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2011-03-31

    There are currently 104 nuclear reactors in 31 states, operated by 51 different utilities. Operation of these reactors generates used fuel assemblies that require storage prior to final disposition. The regulatory framework within the United States (U.S.) allows for the licensing of used nuclear fuel storage facilities for an initial licensing period of up to 40 years with potential for license extensions in 40 years increments. Extended storage, for periods of up to 300 years, is being considered within the U.S. Therefore, there is an emerging need to develop the technical bases to support the licensing for long-term storage. In support of the Research and Development (R&D) activities required to support the technical bases, a comprehensive assessment of the current inventory of used nuclear fuel based upon publicly available resources has been completed that includes the most current projections of used fuel discharges from operating reactors. Negotiations with the nuclear power industry are ongoing concerning the willingness of individual utilities to provide information and material needed to complete the R&D activities required to develop the technical bases for used fuel storage for up to 300 years. This report includes a status of negotiations between DOE and industry in these regards. These negotiations are expected to result in a framework for cooperation between the Department and industry in which industry will provide and specific information on used fuel inventory and the Department will compensate industry for the material required for Research and Development and Testing and Evaluation Facility activities.

  2. On steady-state concentrations of ammonia and molecular hydrogen in the primary circuit of the WWER-1000 reactors

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kamakchi, S.A.

    1997-01-01

    It is shown that the MORAVA-N2 software package describes well the coolant state in the primary circuit of an actual reactor facility with the WWER-1000 during on-load operation. It permits using the package for analysis of process perturbation effect on the coolant composition. Specific feature of ammonia radiation chemistry in the primary circuit of a reactor facility with the WWER-1000, assuring the rates hydrogen concentration in the coolant with ammonia concentration variation in the coolant within wide limits, when reactor operates on power, can be mentioned by way of example, the fact being ascertained in this study

  3. Major alternatives for government policies, organizational structures, and actions in civilian nuclear reactor emergency management in the United States

    International Nuclear Information System (INIS)

    1980-01-01

    The purpose of this report is to identify and assess major alternatives for governmental policies, organizational structures, and actions in civilian nuclear reactor emergency management in the United States. The National Academy of Public Administration agreed to identify and evaluate alternatives for governmental policies, organizational structures, and actions in civilian nuclear reactor emergency management. It agreed to review present policies and practices in civilian nuclear reactor emergency management, to review selected experiences and practices of governmental agencies other than the Nuclear Regulatory Commission, and industries other than the nuclear power industry, and to identify alternatives to the present nuclear emergency system

  4. Spent fuel from RA reactor inspection of state and options for management

    International Nuclear Information System (INIS)

    Aden, V.G.; Bulkin, S. Yu.; Sokolov, A. V.; Matausek, M.V.; Vukadin, Z.

    2001-01-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. The realization of this technology was started in 1999 but due to political and financial difficulties was not completed. In September the year 2000 the work was restarted. Two different ways of RA reactor spent fuel elements preparation for transportation or long-term storage are considered: 'all fuel elements canning without leak-tightness testing' and 'all fuel elements leak-tightness testing'. It is believed that the first option offers several distinct advantages, which can be summarized as: greater reliability in the course of transportation or dry storage. Higher safety for workers. Lower expenditures for non-standard equipment manufacturing. Shorter duration of work. (author)

  5. Parametric study of the primary and secondary systems of the CAREM-25 reactor on steady state

    International Nuclear Information System (INIS)

    Halpert, Silvia; Vazquez, Luis

    2000-01-01

    In the CAREM-25 reactor the primary coolant flows by natural convection that's why the flow is established when the balance between the buoyancy force and friction pressure drop through circuit is obtained. This paper presents a parametric study on primary and secondary systems of the reactor on steady state, for different values of some thermohydraulics parameters: safety factor on friction loss pressure calculations (f), steam generator heat transfer area (A T ) and primary pressure (P P ). The ESCAREM 2.08 thermohydraulic code, which calculates the primary system behavior for steady state conditions, was used for this study. The conclusions of this study are: -) There was a variation of the 15% on the primary coolant flow when the safety factor was changed a 50 %; -) The primary and secondary systems conditions do not change when the power is less than 100 MW; -) Between 100 and 110 MW the decrease of the heat transfer area produces an important change on the secondary systems conditions: the outlet steam generator temperature decrease and there is an important rice in the flow; -) The primary pressure could decrease up to 11.4 MPa without violating turbine requirements. (author)

  6. Qualitative assessment of the value of the Ohio State University TRIGA reactor

    International Nuclear Information System (INIS)

    Binney, S.E.; Johnson, A.G.

    1989-01-01

    The Oregon State University (OSU) TRIGA Reactor (OSTR) is a major regional research, training, and service facility. The OSTR supports a wide variety of organizations at the local, state, regional, national, and international levels. Examples of usage of the OSTR are given in this paper to serve as a basis for assessing the value of the OSTR to its user organizations. It is difficult to assess the value of a facility such as the OSTR quantitatively, primarily because a dollar value cannot be assigned to many of the services that the OSTR performs, e.g., forensic analysis to assist police agencies in criminal cases. Significant qualitative statements can be made, however, to demonstrate the fact that the value of a research reactor facility such as the OSTR substantially outweighs the capital and operating costs of such a facility. Analysis of the data presented above clearly indicates that the value of the OSTR facility is overwhelmingly positive, i.e., the benefits associated with the services provided by the OSTR facility outweigh the cost of providing such services by perhaps as much as an order of magnitude

  7. Solid-state track recorder neutron dosimetry in light water reactor pressure vessel surveillance mockups

    International Nuclear Information System (INIS)

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

    1984-09-01

    Solid-State Track Recorder (SSTR) measurements of neutron-induced fission rates have been made in several pressure vessel mockup facilities as part of the US Nuclear Regulatory Commission's (NRC) Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP). The results of extensive physics-dosimetry measurements made at the Pool Critical Assembly (PCA) at Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN are summarized. Included are 235 U, 238 U, 237 Np and 232 Th fission rates in the PCA 12/13, 8/7, and 4/12 SSC configurations. Additional low power measurements have been made in an engineering mockup at the VENUS critical assembly at CEN-SCK, Mol, Belgium. 237 Np and 238 U fission rates were made at selected locations in the VENUS mockup, which models the in-core and near-core regions of a pressurized water reactor (PWR). Absolute core power measurements were made at VENUS by exposing solid-state track recorders (SSTRs) to polished fuel pellets within in-core fuel pins. 8 references, 4 figures, 10 tables

  8. Stress state variations among the clay and limestone formations of the molasse basin of Northern Switzerland

    International Nuclear Information System (INIS)

    Vietor, Tim; Mueller, Herwig; Frieg, Bernd; Klee, Gerd

    2012-01-01

    Document available in extended abstract form only. Full text of publication follows: The design of geological repositories for radioactive waste responds to the requirements of technical feasibility and long-term safety in the context of a specific geological setting. An important aspect of the geological setting is the primary stress field. To a large extent the stress state controls repository induced effects such as the excavation damage zone and the associated potential changes in the waste isolation properties of the host rock. Therefore the measurement of the stress state receives some attention where the site selection for geological repositories focuses onto relatively weak host rocks such as clay-stones and marly shales that tend to develop a significant excavation damage zone. Measurements of the minimum stress magnitudes in a recently drilled geothermal well in the Molasse Basin of northern Switzerland have yielded a stress profile reaching from 592 m to 1455 m depth. It straddles several rock units and includes the top of the crystalline basement. The sedimentary sequence consists of Marine limestones, shales and marls unconformably covered by Tertiary rocks of the Molasse. In other parts of the basin the evaporitic rocks of the Triassic Muschelkalk formation at the base of the sedimentary layer served as a regional detachment and enabled thin skinned thrusting and the formation of the Jura Fold and Thrust Belt in the Late Miocene. The stress measurements have been performed in the open hole by Mini-frac tests. The method uses a double packer system to isolate a one meter long interval of the borehole that is then pressurized at high injection rates up to the breakdown of the formation. Repeated pressurization of the interval allows to determine the stress that acts on the newly created fracture. The total injected volume during such a test is in the range of a few litres and the size of the fracture that extends from the borehole normal to the minimum

  9. International topical meeting on research reactor fuel management (RRFM) - United States Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) acceptance program: 2007 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.

    2007-01-01

    The Nuclear Weapons Non-proliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by The United States Department of Energy (DOE), in consultation with the Department of State in May 1996, has been extended to expire May 12, 2016, providing an additional 10 years to return fuel to the U.S. This paper provides a brief update on the program, now transferred to the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of nuclear materials (27 countries have participated so far, returning a total of 7620 spent nuclear fuel elements), which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. More than ever before, DOE and reactor operators need to work together to schedule shipments as soon as possible, to optimize shipment efficiency over the remaining years of the program. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible reactor who have not yet participated in the program

  10. Magnets for fusion reactors and plasma physics research: state of the art in the United States

    International Nuclear Information System (INIS)

    Smith, G.E.

    1977-01-01

    The breadth of magnet development in the United States is covered briefly and a few of the difficult technical issues are touched on. Toroidal field coils for tokamaks and superconducting field coils for mirror devices are covered. Parameters of the magnets of various devices are tabulated

  11. Magnets for fusion reactors and plasma physics research: state of the art in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Smith, G. E.

    1977-01-01

    The breadth of magnet development in the United States is covered briefly and a few of the difficult technical issues are touched on. Toroidal field coils for tokamaks and superconducting field coils for mirror devices are covered. Parameters of the magnets of various devices are tabulated. (MHR)

  12. Twenty-ninth annual progress report of the Pennsylvania State University Breazeale Nuclear Reactor, July 1, 1983-June 30, 1984

    International Nuclear Information System (INIS)

    Levine, S.H.; Totenbier, R.E.

    1984-07-01

    The twenty-ninth annual progress report of the operation of the Pennsylvania State University Breazeale Reactor is submitted in accordance with the requirements of Contract DE-AC02-76ER03409 with the United States Department of Energy. This report also provides the University administration with a summary of the operation of the facility for the past year

  13. The Text of the Agreement for the Application of Agency Safeguards to Four United States Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1962-05-24

    The text of the Agreement between the Agency and the United States of America for the application of Agency safeguards to four United States reactor facilities, which was signed on 30 March 1962 and will enter into force on 1 June 1962, is reproduced in this document for the information of all Members of the Agency.

  14. Development of essentialist thinking about religion categories in Northern Ireland (and the United States).

    Science.gov (United States)

    Smyth, Kirsty; Feeney, Aidan; Eidson, R Cole; Coley, John D

    2017-03-01

    Social essentialism, the belief that members of certain social categories share unobservable properties, licenses expectations that those categories are natural and a good basis for inference. A challenge for cognitive developmental theory is to give an account of how children come to develop essentialist beliefs about socially important categories. Previous evidence from Israel suggests that kindergarteners selectively engage in essentialist reasoning about culturally salient (ethnicity) categories, and that this is attenuated among children in integrated schools. In 5 studies (N = 718) we used forced-choice (Study 1) and unconstrained (Studies 2-4) category-based inference tasks, and a questionnaire (Study 5) to study the development of essentialist reasoning about religion categories in Northern Ireland (Studies 1-3 & 5) and the U.S. (Study 4). Results show that, as in Israel, Northern Irish children selectively engage in essentialist reasoning about culturally salient (religion) categories, and that such reasoning is attenuated among children in integrated schools. However, the development trajectory of essentialist thinking and the patterns of attenuation among children attending integrated schools in Northern Ireland differ from the Israeli case. Meta-analysis confirmed this claim and ruled out an alternative explanation of the results based on community diversity. Although the Northern Irish and Israeli case studies illustrate that children develop selective essentialist beliefs about socially important categories, and that these beliefs are impacted by educational context, the differences between them emphasize the importance of historical, cultural, and political context in understanding conceptual development, and suggest that there may be more than one developmental route to social essentialism. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  15. Development of a Risk-Based Performance Assessment Method for Long-Term Cover Systems--Application to the Monticello Mill Tailings Repository

    International Nuclear Information System (INIS)

    HO, CLIFFORD K.; ARNOLD, BILL W.; COCHRAN, JOHN R.; WEBB, STEPHEN W.; TAIRA, RANDAL Y.

    2001-01-01

    A probabilistic, risk-based performance-assessment methodology is being developed to assist designers, regulators, and involved stakeholders in the selection, design, and monitoring of long-term covers for contaminated subsurface sites. This report presents an example of the risk-based performance-assessment method using a repository site in Monticello, Utah. At the Monticello site, a long-term cover system is being used to isolate long-lived uranium mill tailings from the biosphere. Computer models were developed to simulate relevant features, events, and processes that include water flux through the cover, source-term release, vadose-zone transport, saturated-zone transport, gas transport, and exposure pathways. The component models were then integrated into a total-system performance-assessment model, and uncertainty distributions of important input parameters were constructed and sampled in a stochastic Monte Carlo analysis. Multiple realizations were simulated using the integrated model to produce cumulative distribution functions of the performance metrics, which were used to assess cover performance for both present- and long-term future conditions. Performance metrics for this study included the water percolation reaching the uranium mill tailings, radon flux at the surface, groundwater concentrations, and dose. Results of this study can be used to identify engineering and environmental parameters (e.g., liner properties, long-term precipitation, distribution coefficients) that require additional data to reduce uncertainty in the calculations and improve confidence in the model predictions. These results can also be used to evaluate alternative engineering designs and to identify parameters most important to long-term performance

  16. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  17. Fast reactor fuel reprocessing development in the United States: an overview

    International Nuclear Information System (INIS)

    Groenier, W.S.; Burch, W.D.

    1979-01-01

    As a result of the reduced nuclear power demand and the growing concerns over the potential proliferation of sensitive nuclear materials, there has not been a necessity to make immediate decisions regarding near-term reprocessing and breeder reactor commercialization. Programs which formed the basic thrust of nuclear development in the early 1970's have already been adjusted: increased emphasis on problems of radioactive waste management; increased attention to nonproliferation objectives and subsequent reorientation of the overall fuel cycle and breeder programs; increased emphasis on a once-through light-water reactor technology; increased concern for a more detailed knowledge of the uranium resource base; reorientation of the uranium enrichment programs; and exploration of alternative fuel cycles (such as thorium) to minimize the use of plutonium. Nevertheless, major strategic decisions still loom over breeder commercialization, the breeder's requisite demand for reprocessing, and the future role of more proliferation-resistant nuclear technologies. The current program in the United States is organized to provide the necessary technology for the reprocessing of breeder fuels on a timetable that is consistent with the reactor development and demonstration program. Also addressed in this paper are the present day concerns of environmental protection, safety, nuclear material safeguards, and proliferation resistance. It is structured on the well-known Purex processing method but includes new efforts aimed at advanced and alternative fuels. At the present time, the program consists mainly of a generic effort that is planned to progress through an integrated equipment engineering demonstration to an eventual pilot-plant operation. Each of these facilities is viewed as a test bed for advanced and alternative processing steps to address the many significant technical and political issues. 16 figures

  18. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  19. Audit of United States portion of the International Thermonuclear Experimental Reactor project

    International Nuclear Information System (INIS)

    1993-01-01

    Worldwide efforts in fusion energy research are designed to develop fusion power as a safe, environmentally sound, and economically competitive source of energy. The International Thermonuclear Experimental Reactor (ITER) project is a worldwide effort to demonstrate the scientific and technological feasibility of fusion power. The European Community, Japan, the Russian Federation, and the United States are collaborating on ITER, with each of the four parties expected to equally share costs and benefits. Shared costs for the current engineering design phase of the project are estimated at $1 billion in 1989 dollars, excluding certain management and support costs to be absorbed by each partner, with an early estimate of $6 billion, also in 1989 dollars, for construction of the reactor. Engineering design formally began in July 1992, and this phase is in its formative stages. The US had already spent about $100 million since 1987 on ITER conceptual design activities and other preparatory activities in advance of the engineering design phase. Because of its cost significance, the importance of ITER to the US fusion energy program, and the project's unique aspects which may provide a framework for future international endeavors, we initiated an audit of the ITER project. The purpose of the audit was to evaluate management controls over the US portion of the ITER project. Our objectives was to determine whether key front-end controls were in place to ensure that the project could be managed in an efficient and effective manner

  20. State of the reactor vessel surveillance programs in Korea and foreign countries

    International Nuclear Information System (INIS)

    Kim, Jeong Kyu; Hwang, Jong Keun; Park, Keon Woo; Kim, Bum Sik; Jeong, Kyung Hoon

    1996-06-01

    ASTM standards are dominating all over the world in the field of the reactor vessel surveillance program. They are mainly used directly or that the national standards in use correspond quite well with ASTM. According to, however, increasing concerns about the protection of environment and safety of nuclear plant, various approaches to establish and reinforce the national standards are made actively in Europe. In addition, some methods to share the nuclear data by integrating the existing test, analysis procedures and units system are considered. For nuclear plants in Korea, MOST Notice No. 92-20 should be applied for all PWRs after UCN units 3 and 4 since it was promulgated at Dec. 1992. The notice almost reflects the contents of ASTM E 185. But, the notice has much to be desired to provide the technical back-ground for reactor vessel surveillance program because it is not a standard such as ASTM or ASME code but regulation such as CFR or RG. Several Korean Standards are also used in limited area of the surveillance program. Therefore, practical requirements and rules for surveillance program are in accordance with the ASTM and CFR. In this report, the state of application of the standards to the surveillance program in Korea and Europe are reviewed and their national standards re compared with US standards or regulations. Current level and the future prospect of surveillance technology for PWR vessel are discussed at this point of view. 15 tabs., 12 figs., 38 refs. (Author)

  1. The NCSU [North Carolina State Univ.] freon PWR [pressurized water reactor] loop

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. A method for statistical steady state thermal analysis of reactor cores

    International Nuclear Information System (INIS)

    Whetton, P.A.

    1980-01-01

    This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes. Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved. (orig.)

  3. Pellet acceleration studies relating to the refuelling of a steady-state fusion reactor

    International Nuclear Information System (INIS)

    Dimock, D.; Jensen, K.; Jensen, V.O.; Joergensen, L.W.; Pecseli, H.L.; Soerensen, H.; Oester, F.

    1975-11-01

    Several methods for refuelling a steady state-fusion reactor have been proposed, and the pellet method seems advantageous if the pellet can be accelerated to the necessary velocity. A study group was formed to analyze this acceleration problem. Two pellet velocity values were considered: 10 4 m/s and 300 m/s. A pellet velocity of 10 4 m/s may be suitable in the case of a reactor, whereas 300 m/s is believed to be a reasonable velocity at which to perform realistic ablation experiments in the near future. A pneumatic acceleration method was found promising. The pressure is either supplied separately or by evaporation of a part of the pellet. In the latter case, a spark behind the pellet should provide the evoporation and the necessary heating of the driving gas. A preliminary test at room temperature with pellets made of beeswax (the density being ten times that of solid hydrogen, and plastic properties similar to those of solid hydrogen) resulted in a pellet velocity of 100 m/s at a modest value of the energy supplied to the spark. (Auth.)

  4. Small power reactor projects in the United States of America and Canada. Information gathered as a result of invitations from Member States

    International Nuclear Information System (INIS)

    1962-01-01

    As part of its activities in connection with the development of nuclear power, and in response to the resolutions adopted by the General Conference, the Agency has been undertaking a continuing study of the technology and economics of small and medium sized power reactors, particularly with reference to the needs of the less-developed countries. This report summarizes the information gathered on the small power reactor projects in the United States of America and Canada, as a result of the opportunity afforded by these Member States to the Agency. It may be recalled that, at the third regular session of the General Conference, the United States Government offered to provide the Agency with relevant technical and economic data on several small power reactor projects of its Atomic Energy Commission. The Agency accepted the offer and since June 1960 it has sent one or two staff members at approximately six-monthly intervals to follow the development of nine power reactor projects in the United States which represent six different reactor systems. Last year, the Agency issued a report summarizing the information obtained through their visits and study of available published literature. The present document, which should be read in conjunction with that document, brings the information up to date and provides additional information on certain phases of the projects already discussed in the last report. Three more power reactor projects are also dealt with, namely the experimental gas-cooled reactor (EGCR), the high temperature gas-cooled reactor (HTGR) and the Hallam nuclear power facility (HNPF). Early in 1962, the Canadian Government expressed its willingness to make available to the Agency relevant information on the NPD and CANDU projects. The coverage of the NPD reactor is based upon the published information supplied by AECL of Canada and the visit by one of the staff members to the NPD site. The Agency wishes to acknowledge with thanks the co-operation extended

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. State-of-the-art incore detector system provides operational and safety benefits: Example, Hanford N Reactor

    International Nuclear Information System (INIS)

    Toffer, H.

    1988-08-01

    A presentation on the operational and safety benefits that can be derived from a state-of-the-art incore neutron monitoring system has been prepared for the DOE/ANL training course on ''The Potential Safety Impact of New and Emerging Technologies on the Operation of DOE Nuclear Facilities.'' Advanced incore neutron flux monitoring systems have been installed in some commercial reactors and should be considered for any new reactor designs or as backfits to existing plants. The recent installation of such a system at the Hanford N Reactor is used as an example in this presentation. Unfortunately, N Reactor has been placed in a cold standby condition and the full core incore system has not been tested under power conditions. Nevertheless, the evaluations that preceded the installation of the full core system provide interesting insight into the operational and safety benefits that could be expected

  7. The system for diagnostics and monitoring of the IBR-2 reactor state. Data acquisition, accumulation and storage of information

    International Nuclear Information System (INIS)

    Ermilov, V.G.; Ivanov, V.V.; Korolev, V.S.; Pepelyshev, Yu.N.; Semashko, S.V.; Tulaev, A.B.

    2000-01-01

    The architectural decisions for a developed distributed system of the IBR-2 pulsed reactor conditions monitoring are described. The system is intended for measurement of the basic reactor parameters, acquisition, storage and processing of information, the current reactor state monitoring, analysis of reactor parameters for a long time operation period both in on-line, and in off-line modes. The system is constructed in the architecture client-server using DBMS MS SQL Server 7.0 The basic hardware components of the system are measuring workstations and devices, processing and user workstations and the central server. The software of the system consists of the measuring programs, data flows dispatching services, client applications for data processing and visualization, and means for preparing data for subsequent presentation in WWW. The basic results of the first system operation phase and prospect of its development are discussed. (author)

  8. State of the art of the fluidized bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, F.; Vilhena, M.T.M.B. de; Streck, E.; Borges, V.; Johansson, M.

    1987-01-01

    A small and simple nuclear reactors with inherent safety using the fluidized bed concept is under research and study. In this paper a brief study neutronics and thermal hydraulics of this reactor concept is presented. (Author) [pt

  9. Neutronic design of a LEU [low enriched uranium] core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Seshadri, M.D.; Aybar, H.S.; Aldemir, T.

    1987-01-01

    The 10 kw HEU fuelled Ohio State University Reactor (OSURR) will be upgraded to operate at 500 kW with standardized 125 g 235 U LEU U 3 Si 2 fuel plates. An earlier scoping study based on two-dimensional diffusion calculations has identified the potential LEU core configurations for the conversion/upgrade of OSURR using the standardized plates in a 16-plate (+ 2 dummy plates) standard and 10-scoping study is improved for a more precise determination of the excess reactivities and safety rod worths for these potential configurations. Comparison of the results obtained by the improved model to experimental results and to the results of full-core Monte Carlo simulations shows excellent agreement. The results also indicate that the conversion/upgrade of OSURR can be realized with three possible LEU core configurations while maintaining a cold, clean shutdown margin of 1.57-1.91 % Δ k/k, depending on the configuration used. (Author)

  10. Studies on solid-state physics carried out with the Saclay reactor (1962)

    International Nuclear Information System (INIS)

    Herpin, A.

    1962-01-01

    This paper deals only with solid-state physics experiments carried out on outgoing beams: rather than giving a general review of the work performed, if refers to only a few of the most important studies or those nearest completion. These are being made with the experimental beams of the two Saclay reactors EL-2, with a central flux of 10 13 n/cm 2 , and - since 1958 - EL-3, whose central flux is equal ta 10 14 n/cm 2 . The experiments are being carried out by two separate groups of physicists, employing different techniques, namely neutron diffraction using a crystal spectrometer, and inelastic scattering using a time-of-flight spectrometer. (author) [fr

  11. Calculation of the thermal and hydraulic states in rod cluster cores of light-water reactors

    International Nuclear Information System (INIS)

    Teichel, H.

    1977-01-01

    For calculating the three-dimensional steady distribution of the thermal and hydraulic states in rod cluster cores of light-water reactors, the subchannel analysis programs COLA 1 and COLA 2 have been developed. Both programs contain a multitude of competing empirical correlations which may be used by choice. The programs COLA 1 and COLA 2 differ in the calculation method and in the treatment of the boundary condition 'equal pressure at the end of all cooling channels' governing the problem. All parts of the programs are identical. By means of recomputed experiments statements on the accuracy of the results to be expected can be made. In addition, the different suitability of both programs for different experimental conditions are shown. (orig.) [de

  12. Inulinase production in a packed bed reactor by solid state fermentation.

    Science.gov (United States)

    Dilipkumar, M; Rajamohan, N; Rajasimman, M

    2013-07-01

    In this work, production of inulinase was carried out in a packed bed reactor (PBR) under solid state fermentation. Kluyveromyces marxianus var. marxianus was used to produce the inulinase using pressmud as substrate. The parameters like air flow rate, packing density and particle size were optimized using response surface methodology (RSM) to maximize the inulinase production. The optimum conditions for the maximum inulinase production were: air flow rate - 0.82 L/min, packing density - 40 g/L and particle size - 0.0044 mm (mesh - 14/20). At these optimized conditions, the production of inulinase was found to be 300.5 unit/gram of dry substrate (U/gds). Copyright © 2013 Elsevier Ltd. All rights reserved.

  13. Safety of nuclear reactors - Part A - unsteady state temperature history mathematical model

    International Nuclear Information System (INIS)

    El-Shayeb, M.; Yusoff, M.Z.; Boosroh, M.H.; Ideris, F.; Hasmady Abu Hassan, S.; Bondok, A.

    2004-01-01

    A nuclear reactor structure under abnormal operations of near meltdown will be exposed to a tremendous amount of heat flux in addition to the stress field applied under normal operation. Temperature encountered in such case is assumed to be beyond 1000 Celsius degrees. A 2-dimensional mathematical model based on finite difference methods, has been developed for the fire resistance calculation of a concrete-filled square steel column with respect to its temperature history. Effects due to nuclear radiation and mechanical vibrations will be explored in a later future model. The temperature rise in each element can be derived from its heat balance by applying the parabolic unsteady state, partial differential equation and numerical solution into the steel region. Calculation of the temperature of the elementary regions needs to satisfy the symmetry conditions and the relevant material properties. The developed mathematical model is capable to predict the temperature history in the column and on the surface with respect to time. (authors)

  14. Severe transient analysis of the Penn State University Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    Borkowski, J.A.

    1988-08-01

    The Penn State University Advanced Light Water Reactor (PSU ALWR) incorporates various passive and active ultra-safe features, such as continuous online injection and letdown for pressure control, a raised-loop primary system for enhanced natural circulation, a dedicated primary reservoir for enhanced thermal hydraulic control, and a secondary shutdown turbine. Because of the conceptual design basis of the project, the dynamic system modeling was to be performed using a code with a high degree of flexibility. For this reason the modeling has been performed with the Modular Modeling System (MMS). The basic design and normal transients have been performed successfully with MMS. However, the true test of an inherently safe concept lies in its response to more brutal transients. Therefore, such a demonstrative transient is chosen for the PSU ALWR: a turbine trip and reactor scram, concurrent with total loss of offsite ac power. Diesel generators are likewise unavailable. This transient demonstrates the utility of the pressure control system, the shutdown turbine generator, and the enhanced natural circulation of the PSU ALWR. The low flow rates, low pressure drops, and large derivative states encountered in such a transient pose special problems for the modeler and for MMS. The results of the transient analyses indicate excellent performance by the PSU ALWR in terms of inherently safe operation. The primary coolant enters full natural circulation, and removes all decay heat through the steam generators. Further, the steam generators continually supply sufficient steam to the shutdown power system, despite the abrupt changeover to the auxiliary feedwater system. Finally, even with coincident failures in the pressurization system, the primary repressurizes to near-normal values, without overpressurization. No core boiling or uncovery is predicted, and consequently fuel damage is avoided. 17 refs., 19 figs., 4 tabs

  15. Study of advanced fission power reactor development for the United States. Volume I

    International Nuclear Information System (INIS)

    1976-01-01

    This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

  16. State of arts and outlook of research reactor safety management promoted by IAEA

    International Nuclear Information System (INIS)

    Hao Xiaofeng

    2005-01-01

    This paper presents the recent activities of IAEA on the research reactor safety, and the trends in the future. According to the present situation of national research reactors, some suggestions are proposed for the cooperation with IAEA on research reactor safety. (author)

  17. Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

  18. Vibration analysis of primary inlet pipe line during steady state and transient conditions of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Ayazuddin, S.K.; Qureshi, A.A.; Hayat, T.

    1997-11-01

    The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from hold-up tank to the reactor pool of Pakistan Research Reactor-1 (PARR-1). The section of the pipeline from heat exchangers to the valve pit is hanger supported in the pump room and the rest of the section from valve pit to the reactor pool is embedded. The PW-IPL is subjected to steady state and transient vibrations. The reactor pumps, which drive the coolant through various circuits mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of the check valve (water hammer). The ASME Boiler and Pressure Vessel code provides data about the acceptable limits of stresses related to the primary static stress due to steady state vibrations. However, due to complexity in the pipe structure, stresses related to the transient vibrations are neglected in the code. In this report attempt has been made to analyzed both steady state and transient vibrations of PW-IPL of PARR-1. Since, both the steady state and transient vibrations affect the hanger-supported section of the PW-IPL, therefore, it was selected for vibration test measurements. In the analysis vibration data was compared with the allowable limits and estimations of maximum pressure build-up, eflection, natural frequency, tensile and shear load on hanger support, and the ratio of maximum combine stress to the allowable load were made. (author)

  19. Present status of reactor physics in the United States and Japan-II. 1. Deterministic Transport Methods for Reactor Analysis

    International Nuclear Information System (INIS)

    Adams, Marvin L.

    2001-01-01

    We discuss deterministic transport methods used today in neutronic analysis of nuclear reactors. This discussion is not exhaustive; our goal is to provide an overview of the methods that are most widely used for analyzing light water reactors (LWRs) and that (in our opinion) hold the most promise for the future. The current practice of LWR analysis involves the following steps: 1. Evaluate cross sections from measurements and models. 2. Obtain weighted-average cross sections over dozens to hundreds of energy intervals; the result is a 'fine-group' cross-section set. 3. [Optional] Modify the fine-group set: Further collapse it using information specific to your class of reactors and/or alter parameters so that computations better agree with experiments. The result is a 'many-group library'. 4. Perform pin cell transport calculations (usually one-dimensional cylindrical); use the results to collapse the many-group library to a medium-group set, and/or spatially average the cross sections over the pin cells. 5. Perform assembly-level transport calculations with the medium-group set. It is becoming common practice to use essentially exact geometry (no pin cell homogenization). It may soon become common to skip step 4 and use the many-group library. The output is a library of few-group cross sections, spatially averaged over the assembly, parameterized to cover the full range of operating conditions. 6. Perform full-core calculations with few-group diffusion theory that contains significant homogenizations and limited transport corrections. We discuss steps 4, 5, and 6 and focus mainly on step 5. One cannot review a large topic in a short summary without simplifying reality, omitting important details, and neglecting some methods that deserve attention; for this we apologize in advance. (author)

  20. Coal Development in the Northern Great Plains. The Impact on Revenues of State and Local Governments. Agricultural Economic Report No. 394.

    Science.gov (United States)

    Stinson, Thomas F.; Voelker, Stanley W.

    Development of Northern Great Plains coal resources will create new demands for state and local government services. Development will also produce increased government revenues. Special taxes on coal production have been enacted in Montana, North Dakota, and Wyoming in order to ensure that state and local governments receive sufficient revenues to…

  1. The current state of abortion law and practice in Northern Ireland.

    Science.gov (United States)

    Daniels, Pauline; Campbell, Patricia; Clinton, Alison

    This paper reviews current abortion law and practice in Northern Ireland (NI). It explores the origins of NI's abortion law and its complexity in relation to current practice. It reviews issues relating to women seeking terminations in NI and Great Britain and reviews attempts by the Family Planning Association in NI to require the Department of Health, Social Services and Public Safety NI to clarify the current legal basis for termination of pregnancy and to provide guidance for health professionals engaged in this practice. The paper also discusses some of the issues surrounding abortion in NI and seeks to explain why this subject is causing controversy and debate, especially following a judicial review in February and Marie Stopes opening a termination service in Belfast.

  2. State-of-the-art for liquid-level measurements applied to in-vessel coolant level for nuclear reactors

    International Nuclear Information System (INIS)

    Anderson, R.L.

    1980-01-01

    The TMI-2 accident indicated that a direct indication of the liquid level in the reactor vessel would have told the operators that the core was being uncovered. This state-of-the-cost survey covered the following methods: heated thermocouple, differential pressure, ultrasonic, capacitance, microwave, time-domain reflectometry, and externally mounted radiation detectors

  3. TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

    International Nuclear Information System (INIS)

    Chalhoub, E.S.

    1980-09-01

    A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

  4. Analysis of reactor power behaviour using estimation of period for the gain adaptation in a state feedback controller

    International Nuclear Information System (INIS)

    Benitez R, J.S.; Perez C, J.H.; Rivero G, T.

    2008-01-01

    In this paper a novel procedure for power regulation in a TRIGA Mark III nuclear reactor is presented. The control scheme combines state variable feedback with a first order predictor, which is incorporated to speed up the power response of the reactor without exceeding the safety requirement imposed by the reactor period. The simulation results using the proposed control strategy attains different values of steady-state power from different values of initial power in short time, complying at all times with the safety restriction imposed on the reactor period. The predictor, derived from the theory of first order numerical integration, produces very good results during the ascent of power. These results include a fast response and independence of the wide variety of potential operating conditions something not easy and even impossible to obtain with other procedures. By using this control scheme, the reactor period is maintained within safety limits during the start up of the reactor, which is normally the operating condition where an occurrence of a period scram is common. However, the predictor can not be used when the power is reaching the desired power level because the instantaneous power increases far above the desired level. Thus, when the power increases above certain power level, the state feedback gain is set constant to a predefined value. This causes some oscillations that decrease in a few seconds. Afterwards, the power response smoothly approaches, with a small overshoot, the desired power. This constraint on the use of the predictor prevents the unbounded increase of the neutron power. The control law proposed requires all the system's state variables. Since only the neutron power is available, it is necessary the estimation of the non measurable states. The key issue of the existence of a solution to this problem has been previously considered. One of the conclusions is that the point kinetic equations are observable under certain restrictions on

  5. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  6. Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2006-01-01

    As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

  7. State of development of high temperature gas-cooled reactors in foreign countries

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

    Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

  8. Small-angle scattering at a pulsed neutron source: comparison with a steady-state reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borso, C S; Carpenter, J M; Williamson, F S; Holmblad, G L; Mueller, M H; Faber, J Jr; Epperson, J E; Danyluk, S S [Argonne National Lab., IL (USA)

    1982-08-01

    A time-of-flight small-angle diffractometer employing seven tapered collimator elements and a two-dimensional gas proportional counter was successfully utilized to collect small-angle scattering data from a solution sample of the lipid salt cetylpyridinium chloride, C/sub 21/H/sub 38/N/sup +/.Cl/sup -/, at the Argonne National Laboratory prototype pulsed spallation neutron source, ZING-P'. Comparison of the small-angle scattering observed from the same compound at the University of Missouri Research Reactor corroborated the ZING-P' results. The results are used to compare the neutron flux available from the ZING-P' source relative to the well characterized University of Missouri source. Calculations based on experimentally determined parameters indicated the time-averaged rate of detected neutrons at the ZING-P' pulsed spallation source to have been at least 33% higher than the steady-state count rate from the same sample. Differences between time-of-flight techniques and conventional steady-state techniques are discussed.

  9. Calculation of the real states of Ignalina NPP Unit 1 and Unit 2 RBMK-1500 reactors in the verification process of QUABOX/CUBBOX code

    International Nuclear Information System (INIS)

    Bubelis, E.; Pabarcius, R.; Demcenko, M.

    2001-01-01

    Calculations of the main neutron-physical characteristics of RBMK-1500 reactors of Ignalina NPP Unit 1 and Unit 2 were performed, taking real reactor core states as the basis for these calculations. Comparison of the calculation results, obtained using QUABOX/CUBBOX code, with experimental data and the calculation results, obtained using STEPAN code, showed that all the main neutron-physical characteristics of the reactors of Unit 1 and Unit 2 of Ignalina NPP are in the safe deviation range of die analyzed parameters, and that reactors of Ignalina NPP, during the process of the reactor core composition change, are operated in a safe and stable manner. (author)

  10. Observation of the state of the nuclear reactor core by means of non-linear observation algorithms

    International Nuclear Information System (INIS)

    Maciel Palacio, F.E.; Espana, M.D.

    1990-01-01

    A combined, variable-adaptive structure, non-linear observer was designed in order to observe the state of the nuclear reactor core, based on the Absolute Stability Theory. The observer was proved under noise and modelling error conditions. Successful results were obtained in the observation of the states in both cases, showing clear improvement in the observation due to the application of adaptive and variable structure ideas. (Author) [es

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

    International Nuclear Information System (INIS)

    1992-07-01

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

  12. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles

  13. Argon-41 production and evolution at the Oregon State University TRIGA Reactor (OSTR)

    International Nuclear Information System (INIS)

    Anellis, L.G.; Johnson, A.G.; Higginbotham, J.F.

    1988-01-01

    In this study, argon-41 concentrations were measured at various locations within the reactor facility to assess the accuracy of models used to predict argon-41 evolution from the reactor tank, and to determine the relationship between argon gas evolution from the tank and subsequent argon-41 concentrations throughout the reactor room. In particular, argon-41 was measured directly above the reactor tank with the reactor tank lids closed, at other accessible locations on the reactor top with the tank lids both closed and open, and at several locations on the first floor of the reactor room. These measured concentrations were then compared to values calculated using a modified argon-41 production and evolution model for TRIGA reactor tanks and ventilation values applicable to the OSTR facility. The modified model was based in part on earlier TRIGA models for argon-41 production and release, but added features which improved the agreement between predicted and measured values. The approximate dose equivalent rate due to the presence of argon-41 in reactor room air was calculated for several different locations inside the OSTR facility. These dose rates were determined using the argon-41 concentration measured at each specific location, and were subsequently converted to a predicted quarterly dose equivalent for each location based on the reactor's operating history. The predicted quarterly dose equivalent values were then compared to quarterly doses measured by film badges deployed as dose-integrating area radiation monitors at the locations of interest. The results indicate that the modified production and evolution model is able to predict argon-41 concentrations to within a factor of ten when compared to the measured data. Quarterly dose equivalents calculated from the measured argon-41 concentrations and the reactor's operating history seemed consistent with results obtained from the integrating area radiation monitors. Given the argon-41 concentrations measured

  14. State of advanced reactor development in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Reutler, H.

    1988-01-01

    The Federal Republic of Germany is engaged in development work on an advanced light-water reactor that is being designed to achieve a conversion factor of 0.9 on U-Pu fuel. With regard to breeder reactors, most efforts are being concentrated on further improving, with the aid of European partners, the safety standards and economic efficiency of fast breeders. Special efforts are being invested in the development and introduction of small, inherently safe high-temperature reactors

  15. Population fragmentation and inter-ecosystem movements of grizzly bears in Western Canada and the Northern United States

    Science.gov (United States)

    Proctor, M.F.; Paetkau, David; McLellan, B.N.; Stenhouse, G.B.; Kendall, K.C.; Mace, R.D.; Kasworm, W.F.; Servheen, C.; Lausen, C.L.; Gibeau, M.L.; Wakkinen, W.L.; Haroldson, M.A.; Mowat, G.; Apps, C.D.; Ciarniello, L.M.; Barclay, R.M.R.; Boyce, M.S.; Schwartz, C.C.; Strobeck, C.

    2012-01-01

    Population fragmentation compromises population viability, reduces a species ability to respond to climate change, and ultimately may reduce biodiversity. We studied the current state and potential causes of fragmentation in grizzly bears over approximately 1,000,000 km 2 of western Canada, the northern United States (US), and southeast Alaska. We compiled much of our data from projects undertaken with a variety of research objectives including population estimation and trend, landscape fragmentation, habitat selection, vital rates, and response to human development. Our primary analytical techniques stemmed from genetic analysis of 3,134 bears, supplemented with radiotelemetry data from 792 bears. We used 15 locus microsatellite data coupled withmeasures of genetic distance, isolation-by-distance (IBD) analysis, analysis of covariance (ANCOVA), linear multiple regression, multi-factorial correspondence analysis (to identify population divisions or fractures with no a priori assumption of group membership), and population-assignment methods to detect individual migrants between immediately adjacent areas. These data corroborated observations of inter-area movements from our telemetry database. In northern areas, we found a spatial genetic pattern of IBD, although there was evidence of natural fragmentation from the rugged heavily glaciated coast mountains of British Columbia (BC) and the Yukon. These results contrasted with the spatial pattern of fragmentation in more southern parts of their distribution. Near the Canada-US border area, we found extensive fragmentation that corresponded to settled mountain valleys andmajor highways. Genetic distances across developed valleys were elevated relative to those across undeveloped valleys in central and northern BC. In disturbed areas, most inter-area movements detected were made by male bears, with few female migrants identified. North-south movements within mountain ranges (Mts) and across BC Highway 3 were more common

  16. Grassland to shrubland state transitions enhance carbon sequestration in the northern Chihuahuan Desert

    Science.gov (United States)

    M. D. Petrie; S. L. Collins; A. M. Swann; P. L. Ford; M. E. Litvak

    2015-01-01

    The replacement of native C4-dominated grassland by C3-dominated shrubland is considered an ecological state transition where different ecological communities can exist under similar environmental conditions. These state transitions are occurring globally, and may be exacerbated by climate change. One consequence of the global increase in woody vegetation may be...

  17. Conceptual design study of quasi-steady state fusion experimental reactor (FEQ-Q), part 1

    International Nuclear Information System (INIS)

    1985-12-01

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 JER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included; core plasma, reactor structure, reactor core components, magnets. (author)

  18. Identification of reactor failure states using noise methods, and spatial power distribution

    International Nuclear Information System (INIS)

    Vavrin, J.; Blazek, J.

    1981-01-01

    A survey is given of the results achieved. Methodical means and programs were developed for the control computer which may be used in noise diagnostics and in the control of reactor power distribution. Statistical methods of processing the noise components of the signals of measured variables were used for identifying failures of reactors. The method of the synthesis of the neutron flux was used for modelling and evaluating the reactor power distribution. For monitoring and controlling the power distribution a mathematical model of the reactor was constructed suitable for control computers. The uses of noise analysis methods are recommended and directions of further development shown. (J.P.)

  19. State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

    2017-02-15

    A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  20. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

    International Nuclear Information System (INIS)

    March-Leuba, J.; Rey, J.M.

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of 'unexpected' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a 'new and improved' state of the art has emerged recently. (authors). 6 figs., 57 refs., 1 appendix

  1. Overview on advanced nuclear reactors: research and deployment in the United States

    International Nuclear Information System (INIS)

    Sandell, L.; Rohrer, S.

    2004-01-01

    For the United States of America, the electricity requirement is expected to continue to rise at rates of approximately 1.8% over the next few years. This means that some 300,000 MW of additional generating capacity need to be made available by 2025. The Energy Policy Act of 2003 is to minimize this expected future growth of electricity consumption and promote research in favor of a diversified energy mix. As a consequence, the U.S. Senate and the House of Representatives passed legislation on electricity generation, on the promotion of, and research into, specific energy sources, and on energy conservation. Currently, coal-fired power plants contribute the largest share to the overall generating capacity. Considerable additions to the generating capacity have been made in the past ten years in gas-fired plants. In the light of the high present gas prices and market volatilities, the construction of new coal-fired power plants is currently under discussion. 103 out of the 436 nuclear power plants at present in operation worldwide are located in the United States. They represent by far the largest share of emission-free generating capacity in the United States. Considerable capacities have been added over the past few years by, up to now, 99 power increases by 0.4 to 17.8%. The Nuclear Power 2010 Program is a joint initiative by the government and industry seeking to further develop advanced nuclear power plant technologies and elaborate a new licensing procedure for nuclear power plants. The proposed licensing procedure and the Westinghouse AP1000, General Electric ESBWR, and AECL ACR-700 advanced reactor lines are presented. (orig.)

  2. Dual Sarcocystis neurona and Toxoplasma gondii infection in a northern sea otter from Washington state, USA

    Science.gov (United States)

    Lindsay, D.S.; Thomas, N.J.; Rosypal, A.C.; Dubey, J.P.

    2001-01-01

    Dual Sarcocystis neurona and Toxoplasma gondii infection was observed in a Northern sea otter from Washington, USA. The animal was found stranded, convulsed, and died shortly thereafter. Encephalitis caused by both S. neurona and T. gondii was demonstrated in histological sections of brain. Immunohistochemical examination of sections with S. neurona specific antisera demonstrated developmental stages that divided by endopolygeny and produced numerous merozoites. PCR of brain tissue from the sea otter using primer pairs JNB33/JNB54 resulted in amplification of a 1100 bp product. This PCR product was cut in to 884 and 216 bp products by Dra I but was not cut by Hinf I indicating that it was S. neurona [J. Parasitol. 85 (1999) 221]. No PCR product was detected in the brain of a sea otter which had no lesions of encephalitis. Examination of brain sections using T. gondii specific antisera demonstrated tachyzoites and tissue cysts of T. gondii. The lesions induced by T. gondii suggested that the sea otter was suffering from reactivated toxoplasmosis. T. gondii was isolated in mice inoculated with brain tissue. A cat that was fed infected mouse brain tissue excreted T. gondii oocysts which were infective for mice. This is apparently the first report of dual S. neurona and T. gondii in a marine mammal.

  3. Study on small long-life LBE cooled fast reactor with CANDLE burn-up. Part 1. Steady state research

    International Nuclear Information System (INIS)

    Yan, Mingyu; Sekimoto, Hiroshi

    2008-01-01

    Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction. The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead-Bismuth is used as coolant. From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

  5. COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1994-03-01

    The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

  6. HYDROGEOLOGY AND CONCEPTUAL MODEL OF THE KARSTIC COASTAL AQUIFER IN NORTHERN YUCATAN STATE, MEXICO

    Directory of Open Access Journals (Sweden)

    Miguel J Villasuso-Pino

    2011-04-01

    Full Text Available The coastal zone of northern Yucatan Peninsula (YP is mainly constituted by Tertiary limestones, covered by Pleistocen limestones, where there exist swamps and estuary systems, locally called “rías”, with mouths connecting them to the sea and hence being a way for an important amount of groundwater to discharge, like in Ría Lagartos and Celestún. These limestones have karstic layers located at depths from 8 to 16 meters below terrain surface.  It is in these layers where groundwater mainly flows toward coast, passing below the sand dune and discharging in the sea in the form of submarine springs which in many cases manifest themselves on the marine surface depending on the hydraulic or piezometric fresh water head. The width of the superficial limestone within this coastal fringe, called “caliche”, varies from 5 to 10 kilometers in the study zone (Chuburna-Progreso-Chicxulub.  Its permeability is extremely low, so it constitutes a confining layer that impedes superficial waters to percolate toward groundwater.  The hydraulic head of the groundwater below this confining layer is over the mean sea level and also over the swamp water level, coastal lagoons and estuaries. There are two important hydrological phenomena that occur in this coastal fringe: 1 There is no recharge to the aquifer (groundwater due to limestone rock outcrops is impermeable or semipermeable; and 2 groundwater pressure is not lost, nor saline interfase is rised if the superficial layer is broken.  The groundwater pollution vulnerability within this coastal fringe is less than that for the superficial saline waters of swamps and estuaries, because of caliche’s low intrinsic permeability that impedes percolation.

  7. Nuclear reactors built, being built, or planned in the United States as of June 30, 1982

    International Nuclear Information System (INIS)

    Goulden, A.M.

    1982-11-01

    This semiannual compilation provides current information about facilities for domestic use or export which are capable of sustaining a nuclear chain reaction. Civilian, production, and military reactors are listed, as are reactors for export and critical assembly facilities. Information given includes location, owner, principal nuclear contractor, type, power rating, docket number, and start-up and shutdown dates

  8. State of the art of nuclear facilities with organic cooled reactors

    International Nuclear Information System (INIS)

    Brede, O.

    1984-01-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined. (author)

  9. Selected power reactor projects in Canada and the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-11-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

  10. Study of advanced fission power reactor development for the United States. Volume II

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

  11. Selected power reactor projects in Canada and the United States of America

    International Nuclear Information System (INIS)

    1964-01-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

  12. The behaviour of impurities in a steady-state DT gas-blanket reactor

    International Nuclear Information System (INIS)

    Markvoort, J.A.

    1975-11-01

    A four-fluid model of a cylindrical steady-state DT gas-blanket reactor is analysed. The four fluids are electrons, deuterium-tritium, helium and a high -Z impurity. The behaviour of the plasma is described by the multifluid MHD-equations which are numerically solved with the aid of a Runge Kutta method. Whether impurities tend to concentrate on the axis is found to depend on how, in the collision term, the Nernst effect is taken into account. In order to show the influence of the Nernst terms arising from electron-ion collisions and the Nernst terms due to ion-ion collisions separately, the thermal force is dealt with in two ways. In model A, only the contribution from electron-ion collisions was considered. The computer calculations show that the impurities have their maximum concentration on the axis. A theoretical analysis explains this result. In model B, which is more realistic, these ion-ion collisions are included. The computer calculations as well as the theoretical analysis show that the influence of the thermoforce due to ion-ion collisions on the density profiles dominates over the force due to electron collisions, and lead to a minimum in the impurity density on the axis. As in model A, the analytical analysis yields relationships between the various density profiles and the temperature profile

  13. HEATHYD, Steady-State Thermal Hydraulic Analysis of Low-Enriched U Fuel Reactor

    International Nuclear Information System (INIS)

    NABBI, R.

    1989-01-01

    1 - Description of program or function: HEATHYD is a code for the steady-state heat transfer calculation of research nuclear reactors with forced convection. It models heat transfer and coolant flow for assemblies of parallel fuel plates of MTR type with any axial power distribution. The thermodynamic model accounts for single phase cooling and sub- cooled boiling condition using the transition criterion of Bergeles-Rosenow. In addition to the calculation of the channel flow velocities and coolant pressure drops, HEATHYD calculates axial distribution of the coolant and clad-surface temperatures. Safety margins to the critical heat flux as a result of burnout condition or flow instability are determined. 2 - Method of solution: Applying the finite difference method, HEATHYD solves the equations of heat conduction and heat transfer to the coolant. For the physical properties of the coolant as a function of the coolant temperature polynomials of degree 6 are used. Depending on the coolant condition, different correlations for the heat transfer coefficient can be applied. The analysis of the critical cooling conditions resulting in burnout or flow instability, is performed according to the correlations developed by Mirshak/ Labuntsov and Forgan/Whittle

  14. Current drive studies for the ARIES steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Ehst, D.A.; Mandrekas, J.

    1994-01-01

    Steady-state plasma operating scenarios are designed for three versions of the ARIES reactor, using non-inductive current drive techniques that have an established database. R.f. waves, including fast and lower hybrid waves, are the reference drivers for the D-T burning ARIES-I and ARIES-II/IV, while neutral beam injection is employed for ARIES-III which burns D- 3 He. Plasma equilibria with a high bootstrap-current component have been used, in order to minimize the recirculating power fraction and cost of electricity. To maintain plasma stability, the driven current profile has been aligned with that of equilibrium by proper choices of the plasma profiles and power launch parameters. Except for ARIES-III, the current-drive power requirements and the relevant technology developments are found to be quite reasonable. The wave-power spectrum and launch requirements are also considered achievable with a modest development effort. Issues such as an improved database for fast-wave current drive, lower-hybrid power coupling to the plasma edge, profile control in the plasma core, and access to the design point of operation remain to be addressed. ((orig.))

  15. Performance analysis of the intense slow-positron beam at the NC State University PULSTAR reactor

    International Nuclear Information System (INIS)

    Moxom, J.; Hathaway, A.G.; Bodnaruk, E.W.; Hawari, A.I.; Xu, J.

    2007-01-01

    An intense positron beam, for application in nanophase characterization, is now under construction at the 1 MW PULSTAR nuclear reactor at North Carolina State University (NCSU). A tungsten converter/moderator is used, allowing positrons to be emitted from the surface with energies of a few electron volts. These slow positrons will be extracted from the moderator and formed into a beam by electrostatic lenses and then injected into a solenoidal magnetic field for transport to one of three experimental stations, via a beam switch. To optimize the performance of the beam and to predict the slow-positron intensity, a series of simulations were performed. A specialized Monte-Carlo routine was integrated into the charged-particle transport calculations to allow accounting for the probabilities of positron re-emission and backscattering from multiple-bank moderator/converter configurations. The results indicate that either a two-bank or a four-bank tungsten moderator/converter system is preferred for the final beam design. The predicted slow-positron beam intensities range from nearly 7x10 8 to 9x10 8 e + /s for the two-bank and the four-bank systems, respectively

  16. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R.

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively

  17. Economic impacts on the United States of siting decisions for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Peerenboom, J.P.; Wolsko, T.D.; Hanson, M.E.

    1997-01-01

    This paper presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively

  18. Burst wait time simulation of CALIBAN reactor at delayed super-critical state

    International Nuclear Information System (INIS)

    Humbert, P.; Authier, N.; Richard, B.; Grivot, P.; Casoli, P.

    2012-01-01

    In the past, the super prompt critical wait time probability distribution was measured on CALIBAN fast burst reactor [4]. Afterwards, these experiments were simulated with a very good agreement by solving the non-extinction probability equation [5]. Recently, the burst wait time probability distribution has been measured at CEA-Valduc on CALIBAN at different delayed super-critical states [6]. However, in the delayed super-critical case the non-extinction probability does not give access to the wait time distribution. In this case it is necessary to compute the time dependent evolution of the full neutron count number probability distribution. In this paper we present the point model deterministic method used to calculate the probability distribution of the wait time before a prescribed count level taking into account prompt neutrons and delayed neutron precursors. This method is based on the solution of the time dependent adjoint Kolmogorov master equations for the number of detections using the generating function methodology [8,9,10] and inverse discrete Fourier transforms. The obtained results are then compared to the measurements and Monte-Carlo calculations based on the algorithm presented in [7]. (authors)

  19. Burst wait time simulation of CALIBAN reactor at delayed super-critical state

    Energy Technology Data Exchange (ETDEWEB)

    Humbert, P. [Commissariat a l' Energie Atomique CEA, Centre de Bruyeres-le-Chatel, 91297 Arpajon (France); Authier, N.; Richard, B.; Grivot, P.; Casoli, P. [Commissariat a l' Energie Atomique CEA, Centre de Valduc, 21120 Is-sur-Tille (France)

    2012-07-01

    In the past, the super prompt critical wait time probability distribution was measured on CALIBAN fast burst reactor [4]. Afterwards, these experiments were simulated with a very good agreement by solving the non-extinction probability equation [5]. Recently, the burst wait time probability distribution has been measured at CEA-Valduc on CALIBAN at different delayed super-critical states [6]. However, in the delayed super-critical case the non-extinction probability does not give access to the wait time distribution. In this case it is necessary to compute the time dependent evolution of the full neutron count number probability distribution. In this paper we present the point model deterministic method used to calculate the probability distribution of the wait time before a prescribed count level taking into account prompt neutrons and delayed neutron precursors. This method is based on the solution of the time dependent adjoint Kolmogorov master equations for the number of detections using the generating function methodology [8,9,10] and inverse discrete Fourier transforms. The obtained results are then compared to the measurements and Monte-Carlo calculations based on the algorithm presented in [7]. (authors)

  20. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R. [and others

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  1. State of the art of toshiba maintenance techniques for reactor internals

    International Nuclear Information System (INIS)

    Maekawa, Osamu; Hattori, Yasuhiro; Sudo, Akira

    2002-01-01

    As the number of aged plants increases, maintaining the integrity of the reactor pressure vessel and reactor internals in aged plants has become an essential issue to ensure continued stable operation and achieve higher plant operability. A major issue with regard to reactor internals is stress corrosion cracks (SCCs). Laser-applying techniques have many features suitable for preventive maintenance work on reactor internals. Toshiba has developed various laser-applying preventive maintenance techniques and accumulated considerable field experience utilizing these techniques in various aged plants. Moreover, in view of the importance of confirming the soundness of reactor internals in aged plants, Toshiba has developed and applied sophisticated nondestructive testing techniques for this purpose. (author)

  2. Changing concepts of geologic structure and the problem of siting nuclear reactors: examples from Washington State

    International Nuclear Information System (INIS)

    Tabor, R.W.

    1986-01-01

    The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alignment might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes - both concepts little-considered during initial site selection - may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting

  3. Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

    International Nuclear Information System (INIS)

    Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia

    2017-01-01

    MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

  4. Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

    Energy Technology Data Exchange (ETDEWEB)

    Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail: ghnfernandes@gmail.com, E-mail: marc5663@gmail.com, E-mail: athos1495@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2017-07-01

    MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

  5. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    International Nuclear Information System (INIS)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-01-01

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  6. Most probable mixing state of aerosols in Delhi NCR, northern India

    Science.gov (United States)

    Srivastava, Parul; Dey, Sagnik; Srivastava, Atul Kumar; Singh, Sachchidanand; Tiwari, Suresh

    2018-02-01

    Unknown mixing state is one of the major sources of uncertainty in estimating aerosol direct radiative forcing (DRF). Aerosol DRF in India is usually reported for external mixing and any deviation from this would lead to high bias and error. Limited information on aerosol composition hinders in resolving this issue in India. Here we use two years of aerosol chemical composition data measured at megacity Delhi to examine the most probable aerosol mixing state by comparing the simulated clear-sky downward surface flux with the measured flux. We consider external, internal, and four combinations of core-shell (black carbon, BC over dust; water-soluble, WS over dust; WS over water-insoluble, WINS and BC over WINS) mixing. Our analysis reveals that choice of external mixing (usually considered in satellite retrievals and climate models) seems reasonable in Delhi only in the pre-monsoon (Mar-Jun) season. During the winter (Dec-Feb) and monsoon (Jul-Sep) seasons, 'WS coating over dust' externally mixed with BC and WINS appears to be the most probable mixing state; while 'WS coating over WINS' externally mixed with BC and dust seems to be the most probable mixing state in the post-monsoon (Oct-Nov) season. Mean seasonal TOA (surface) aerosol DRF for the most probable mixing states are 4.4 ± 3.9 (- 25.9 ± 3.9), - 16.3 ± 5.7 (- 42.4 ± 10.5), 13.6 ± 11.4 (- 76.6 ± 16.6) and - 5.4 ± 7.7 (- 80.0 ± 7.2) W m- 2 respectively in the pre-monsoon, monsoon, post-monsoon and winter seasons. Our results highlight the importance of realistic mixing state treatment in estimating aerosol DRF to aid in policy making to combat climate change.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  8. Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

    International Nuclear Information System (INIS)

    Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

    2015-01-01

    The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

  9. Outage performance improvement by state of the art reactor stud tensioning

    International Nuclear Information System (INIS)

    Oehler, Horst Werner; Vervliet, Herman

    2006-01-01

    Actual methods of reactor closing, i.e. cover to vessel sealing, is based on the creation of an equal load to the sealing circumference by tensioning all reactor studs with an equal force. This method ensures leak tightness through equal compression of the reactor seal in normal circumstances and is largely applied for all types of reactors throughout many generations and designs of nuclear power stations. The tension generated in each reactor stud is controlled indirectly by measuring the reactor stud elongation while under stress. Most studs are designed to measure this elongation easily by conventional or more advanced systems (from individual clock gauge to integrated digital transmission to a computer screen). It is this elongation value, prescribed by the reactor vessel/cover manufacturer which must be respected and demonstrated during all reactor closing operations, weather they take place for initial hydro testing, refuelling operations or periodical hydraulic tests of the primary circuit. Closing (and re-opening) of reactor vessels has become a routine operation as it is required for fuel reloading of the reactor core. This operation is performed on all PWR and BWR type of reactors with a large variety of tooling. As most of the utilities have implemented maintenance optimisation programs, the refuelling outage is reduced to a sequence of activities that allow quick and efficient refuelling of the core. The performance and efficiency of instrumentation and tooling deployed during these essential activities are of the utmost importance to minimise the critical path of the refuelling outage. Today, in support of outage performance, many utilities have invested in new and refurbished tooling to allow quick and efficient opening and closing of the reactor vessel. The features and properties of the most performing multi stud tensioning machines currently in service in nuclear power stations world wide (Africa, Europe, Asia and USA) are presented in the paper

  10. Fishing activity in Northern Rio de Janeiro State (Brazil and its relation with small cetaceans

    Directory of Open Access Journals (Sweden)

    Ana Paula Madeira Di Beneditto

    1998-01-01

    Full Text Available Research on fishing activity at Atafona village, in Northern Rio de Janeiro, Brazil (21°35'S, was carried out between 1987-96 for the purpose of relating it to the accidental capture of small cetaceans and of estimating the relationship between fishing activity and the diet of small cetaceans. Data on fishing operations were obtained at the cold storage plants management, from interviews with fishermen and personal observations. The most representative fishing resources were Xyphopenaeus kroyeri, Micropogonias furnieri, Carcharhinus plumbeus, C. acronotus,and Rhizoprionodon porosus. Gillnets are responsible for the accidental capture of small cetaceans in the region, mainly Pontoporia blainvillei and Sotalia fluviatilis (marine form. Four types of gillnets that are used on the region ("minjuada", "sarda", "caçoá" and "pescadinha" were dangerous to these species because they are placed in their preferred habitat. There is no competition between fishermen and small cetaceans due to the selection in the capture of commercialized fishesInvestigação sobre a atividade pesqueira na localidade de Atafona, Norte do Rio de Janeiro, Brasil (21º25`S, foi conduzida entre 1987-96 com o objetivo de relacioná-la com a captura acidental e a dieta dos pequenos cetáceos. Dados sobre as operações pesqueiras foram obtidos na administração dos entrepostos de pesca, através de entrevistas com pescadores e observações pessoais. Os recursos pesqueiros mais representativos foram Xyphopenaeus kroyeri, Micropogonias furnieri, Carcharhinus plumbeus, C. acronotus, and Rhizoprionodon porosus. As redes de espera são responsáveis pela captura acidental de pequenos cetáceos na região, principalmente de Pontoporia blainvillei e Sotalia fluviatilis (forma marinha. Quatro tipos de redes de espera que são usadas na região ("minjuada", "sarda", "caçoá" and "pescadinha" foram mais perigosas para essas espécies pois são colocadas no seu hábitat preferencial

  11. Food Defense Practices of School Districts in Northern U.S. States

    Science.gov (United States)

    Klitzke, Carol J.

    2013-01-01

    This study assessed implementation of food defense practices in public schools in Montana, Wyoming, South Dakota, North Dakota, Iowa, Minnesota, and Wisconsin. The first phase involved a qualitative multi-site case study: one-day visits were made to five school districts in the states of Iowa, South Dakota, Minnesota, and Wisconsin. A principal,…

  12. Present state and future prospect of development of high temperature gas-cooled reactors in Japan

    International Nuclear Information System (INIS)

    Sanokawa, Konomo

    1994-01-01

    High temperature gas-cooled reactors can supply the heat of about 1000degC, and the high efficiency and the high rate of heat utilization can be attained. Also they have the features of excellent inherent safety, the easiness of operation, the high burnup of fuel and so on. The heat utilization of atomic energy in addition to electric power generation is very important in view of the protection of global environment and the diversification of energy supply. Japan Atomic Energy Research Institute has advanced the construction of the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output, aiming at attaining the criticality in 1998. The progress of the development of a high temperature gas-cooled reactor is described. For 18 years, the design study of the reactor was advanced together with the research and development of the reactor physics, fuel and materials, high temperature machinery and equipment and others, and the decision of the design standard and the development of computation codes. The main specification and the construction schedule are shown. The reactor building was almost completed, and the reactor containment vessel was installed. The plan of the research and development by using the HTTR is investigated. (K.I.)

  13. Analyses of the transportation of spent research reactor fuel in the United States

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    The Transportation Technology Center at Sandia National Laboratories has analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy (DOE) Office of Defense Programs. This effort represents the first comprehensive analytical evaluation of the risks of transporting high-, medium-, and low-enriched uranium spent research reactor fuel by both sea and land. Two separate shipment programs have been analyzed: the shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). In order to perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

  14. Effects of land use/cover change and harvests on forest carbon dynamics in northern states of the United States from remote sensing and inventory data: 1992-2001

    Science.gov (United States)

    Daolan Zheng; Linda S. Heath; Mark J. Ducey; James E. Smith

    2011-01-01

    We examined spatial patterns of changes in forest area and nonsoil carbon (C) dynamics affected by land use/cover change (LUC) and harvests in 24 northern states of the United States using an integrated methodology combining remote sensing and ground inventory data between 1992 and 2001. We used the Retrofit Change Product from the Multi-Resolution Land Characteristics...

  15. Nuclear reactors built, being built, or planned in the United States

    International Nuclear Information System (INIS)

    Goulden, A.M.

    1983-08-01

    This semiannual compilation provides current information about facilities for domestic use or export which are capable of sustaining a nuclear chain reaction. Civilian, production, and military reactors are listed, as are reactors for export and critical assembly facilities. Information given includes location, owner, principal nuclear contractor, type, power rating, docket number, and start-up and shutdown dates. Nuclear Reactors Built, Being Built, or Planned is also available on standing order (PB83-903000) through a deposit account with the National Technical Information Service, Springfield, VA 22161

  16. Development of the heavy-water organic-cooled reactor. Status report from the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Trilling, C A [Atomics International, Division of North American Aviation, Inc., Canoga Park, CA (United States)

    1967-01-01

    In late 1964 the United States Atomic Energy Commission decided to undertake the development of the heavy-water-moderated nuclear power reactor as part of its overall programme for the development of advanced converter reactors. The inclusion of the heavy-water reactor concept was based on its indicated potential for achieving: efficient utilization of available fuel resources; generation of low cost electric power; feasibility of scale-up to very large single unit plant sizes for the dual purpose of generating power and desalting sea water. The excellent neutron economy inherent in heavy-water moderation allows a significant increase in the amount of power which can be generated from a given amount of ore. If one takes into account the amount of uranium required not only for burn-up but also to inventory new reactors in a rapidly expanding nuclear economy, heavy-water reactors show the potential of extracting one and a half to two times more power from the ore mined than light-water reactors. Such an improvement in dynamic fuel utilization will postpone the depletion of low cost uranium ore reserves, providing more time for the discovery of new ore resources and the development of economic fast breeder reactors. The excellent neutron economy of the heavy-water reactor also allows the achievement of appreciable burn-up with low enrichment fuel, with consequent low fuel cycle costs and therefore low energy generation costs. These low fuel cycle costs make the economics of this type of reactor rather insensitive to rising ore costs. They also make the concept well suited for the most economic production of the large quantities of heat required for water desalination. The use of individual pressure tubes for circulating the coolant through the reactor vessel lends itself to the development of a modular type design, which can be scaled up to very large single unit plant sizes by simply increasing the number of identical pressure tube modules and the number of coolant

  17. Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z., E-mail: raissaomarques@gmail.com, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

  18. Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

    International Nuclear Information System (INIS)

    Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z.

    2017-01-01

    Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

  19. POTENTIAL IMPACTS OF GM WHEAT ON UNITED STATES AND NORTHERN PLAINS WHEAT TRADE

    OpenAIRE

    Taylor, Richard D.; DeVuyst, Eric A.; Koo, Won W.

    2003-01-01

    The potential introduction of genetically modified (GM) wheat has both supporters and opponents waging battle in the popular press and scholarly research. Supporters highlight the benefits to producers, while the opponents highlight the unknown safety factors for consumers. The topic is very important to the United States, as a large portion of the wheat production is exported overseas. Consumer groups in some countries are resisting GM wheat. This study utilizes a spatial equilibrium model t...

  20. Transportation infrastructure upgrades in the South: A compilation of state plans for construction near nuclear reactor sites

    International Nuclear Information System (INIS)

    1992-03-01

    There are currently 27 nuclear reactor sites located in the southern region. In many instances, the most practicable modes of transportation of spent nuclear fuel from these sites we through the use of highway and rail systems. These two transportation modes have important differences that affect their applicability; chief among these, perhaps, is the fact that while highway systems are publicly owned and maintained rail lines are owned by private entities. For this reason, track condition and maintenance, usage rates and other aspects of rail transport can vary widely. This report reviews southern state, department plans for infrastructure upgrades in the vicinity of nuclear reactor sites. This report includes a summary of planned modifications to bridges, access highways, and rail spurs (where applicable) over the next five years. The information contained herein was gathered from interviews with officials within state departments of transportation. With few exceptions, the contact person was an official within the departmental planning division

  1. Nuclear reactors: built, being built, or planned in the United States as of Dec 31, 1979

    International Nuclear Information System (INIS)

    1980-07-01

    Information is tabulated in nuclear reactor and critical assembly facilities in operation, shut down, under construction, or planned. The data include name, owner, location, type, power, and startup date

  2. Nuclear reactors built, being built, or planned in the United States as of June 30, 1980

    International Nuclear Information System (INIS)

    Goulden, A.M.

    1980-12-01

    Information is tabulated on nuclear reactor and critical assembly facilities in operation, shut down, under construction, or planned. The data included name, owner, location, type, power, and startup date

  3. Design of a reactor inlet temperature controller for EBR-2 using state feedback

    International Nuclear Information System (INIS)

    Vilim, R.B.; Planchon, H.P.

    1990-01-01

    A new reactor inlet temperature controller for pool type liquid-metal reactors has been developed and will be tested in EBR-II. The controller makes use of modern control techniques to take into account stratification and mixing in the cold pool during normal operation. Secondary flowrate is varied so that the reactor inlet temperature tracks a setpoint while reactor outlet temperature, primary flowrate and secondary cold leg temperature are treated as exogenous disturbances and are free to vary. A disturbance rejection technique minimizes the effect of these disturbances on inlet temperature. A linear quadratic regulator improves inlet temperature response. Tests in EBR-II will provide experimental data for assessing the performance improvements that modern control can produce over the existing EBR-II analog inlet temperature controller. 10 refs., 8 figs

  4. Status of Liquid Metal Fast Reactor Development in the United States of America, March 1987

    International Nuclear Information System (INIS)

    Horton, K.E.

    1987-01-01

    In order to meet the objective to develop and demonstrate economically competitive reactor designs and associated fuel cycles early in the next century, the U.S. program has become more focused. Two innovative reactor designs supported by the metal-fueled Integral Fast Reactor program are being directed at fulfilling a series of advanced reactor goals. The supporting technology programs and facilities are being refocused to support the overall goals. International collaboration is being broadened to provide the two-way support across the spectrum of plant projects and the fuel cycle. This program is intended to maintain the technology base into the time period (mid-1990s) when a private sector demonstration could be initiated. (author)

  5. Status of fast breeder reactor development in the United States of America - April 1985

    International Nuclear Information System (INIS)

    Horton, K.E.

    1986-01-01

    In order to provide a continuum of development for liquid metal reactors, the U.S. program has been reshaped into two portions -- advanced converter reactor technology including the Liquid Metal Reactor for the near and intermediate term, and the Liquid Metal Fast Breeder Reactor for deployment in the twenty-first century. The focused research and development program is directed at innovative means to improve economics, provide inherent safety, and to meet the needs of the ultimate user, the utilities. Work in the fuels and materials, and reprocessing areas is being continued to support eventual deployment. A major factor in successful deployment will be the effectiveness of international collaboration in reducing costs and duplication of efforts

  6. Reflections on the introduction of fast breeder reactors in the DeBeNeLux states

    International Nuclear Information System (INIS)

    Schroeder, R.; Wagner, J.

    1975-06-01

    This report gives a survey of the impact of introducing sodium-cooled fast breeder reactors in the Federal Republic of Germany and the BeNeLux countries (DeBeNeLux region). The supply situation with respect to electric and thermal energy is studied in particular, together with aspects of economy and environmental impact. The potential and consequences of a breeder economy, the present status and future r+d work are discussed. In addition to sodium-cooled fast breeder reactors with oxide or carbide fuel, alternative solutions are touched: 1) light water and high temperature reactors, 2) helium-cooled fast breeder reactors, 3) geothermal energy, solar energy and fusion energy. (orig.) [de

  7. Space reactor/organic Rankine conversion - A near-term state-of-the-art solution

    Science.gov (United States)

    Niggemann, R. E.; Lacey, D.

    The use of demonstrated reactor technology with organic Rankine cycle (ORC) power conversion can provide a low cost, minimal risk approach to reactor-powered electrical generation systems in the near term. Several reactor technologies, including zirconium hydride, EBR-II and LMFBR, have demonstrated long life and suitability for space application at the operating temperature required by an efficient ORC engine. While this approach would not replace the high temperature space reactor systems presently under development, it could be available in a nearer time frame at a low and predictable cost, allowing some missions requiring high power levels to be flown prior to the availability of advanced systems with lower specific mass. Although this system has relatively high efficiency, the heat rejection temperature is low, requiring a large radiator on the order of 3.4 sq m/kWe. Therefore, a deployable heat pipe radiator configuration will be required.

  8. Analyses of the transportation of spent research reactor fuel in the United States

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

  9. Safety Evaluation Report related to the renewal of the operating license for the research reactor at Michigan State University (Docket No. 50-294)

    International Nuclear Information System (INIS)

    1984-08-01

    This Safety Evaluation Report for the application filed by the Michigan State University (MSU) for a renewal of operating license number R-114 to continue to operate the TRIGA Mark I research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Michigan State University and is located on the campus of Michigan State University in East Lansing, Ingham County, Michigan. The staff concludes that the TRIGA reactor facility can continue to be operated by MSU without endangering the health and safety of the public

  10. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

    1994-05-01

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  12. Method for calculating the steady-state distribution of tritium in a molten-salt breeder reactor plant

    International Nuclear Information System (INIS)

    Briggs, R.B.; Nestor, C.W.

    1975-04-01

    Tritium is produced in molten salt reactors primarily by fissioning of uranium and absorption of neutrons by the constituents of the fuel carrier salt. At the operating temperature of a large power reactor, tritium is expected to diffuse from the primary system through pipe and vessel walls to the surroundings and through heat exchanger tubes into the secondary system which contains a coolant salt. Some tritium will pass from the secondary system into the steam power system. This report describes a method for calculating the steady state distribution of tritium in a molten salt reactor plant and a computer program for making the calculations. The method takes into account the effects of various processes for removing tritium, the addition of hydrogen or hydrogenous compounds to the primary and secondary systems, and the chemistry of uranium in the fuel salt. Sample calculations indicate that 30 percent or more of the tritium might reach the steam system in a large power reactor unless special measures are taken to confine the tritium. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. Development of the selection system in northern hardwood forests of the Lake States: an 80-year silviculture research legacy

    Science.gov (United States)

    Christel Kern; Gus Erdmann; Laura Kenefic; Brian Palik; Terry. Strong

    2014-01-01

    The northern hardwood research program at the Dukes Experimental Forest in Michigan and Argonne Experimental Forest in Wisconsin has been adapting to changing management and social objectives for more than 80 years. In 1926, the first northern hardwood silviculture study was established in old-growth stands at the Dukes Experimental Forest. In response to social...

  15. Radiologic states of the WWR-S Bucharest Reactor following definitive shutdown

    International Nuclear Information System (INIS)

    Garlea, C.; Kelerman, C.; Mocioiu, D.; Garlea, I.

    2001-01-01

    The definitive shutdown of a reactor raises problems related to the management of the radioactive inventory. To define the radioactive inventory contained in the burned nuclear fuel and in the neutron activated structural materials computation methods are to be used. Besides the radioactive inventory contained in the main block of the reactor, the one due to the primary circuit contaminated mainly with fission products and corrosion products activated in the reactor core, transported and deposed on the components of the cooling primary circuit should be added. Also another component of the radioactive inventory intervenes, namely, the one due to the contamination of the technological rooms used for various operations the nuclear activities (hot cells, pump room, reactor hall, passage ways to the hot cells and for radioactive source, radioisotope and radioactive waste transport). The activities which made used of the neutron and gamma fluxes for radioisotope production, materials irradiation, research, component testing, resulted in radioactive waste, technological or accidental contaminations of the technological rooms of the reactor. Inspections and current repair interventions resulted also in radioactive waste an contaminations. Consequently systematic measurements with qualified equipment dedicated to alpha, beta, gamma contamination measurements as well as to dose rates determinations for the personnel exposed are necessary. Irrespective of the duration of the reactor conservation or shutdown, the radiologic monitoring should continue. This work presents the results obtained by the research group 'Restoration of Nuclear Sites', working with the IFIN-HH, regarding both the radioactive inventory calculation and measurements of contamination of technological rooms and environment in the reactor vicinity

  16. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

    2013-09-15

    Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

  17. Two new endangered species of Anomaloglossus (Anura: Aromobatidae) from Roraima State, northern Brazil.

    Science.gov (United States)

    Fouquet, Antoine; Souza, Sergio Marques; Nunes, Pedro M Sales; Kok, Philippe J R; Curcio, Felipe Franco; De Carvalho, Celso Morato; Grant, Taran; Rodrigues, Miguel Trefaut

    2015-03-05

    We describe two new species of Anomaloglossus from Roraima State, Brazil, that are likely endemic to single mountains currently isolated among lowland forest and savanna ecosystems. The first species, Anomaloglossus tepequem sp. nov. was collected in 1986 and 1992 along a single stream at >500 m elevation on a tepui-like mountain named Tepequém, but was not detected during recent investigations. It is mainly diagnosed from other Anomaloglossus species by its well developed foot webbing, immaculate cream abdomen colouration and small body size (males: 18.2-20.1 mm, females: 21.7-24.5). The second species, Anomaloglossus apiau sp. nov. was found along several streams between 500 and 1400 m elevation on Serra do Apiaú, and is mainly diagnosed from congeners by its weakly webbed feet, males with swollen third finger and ventrolateral stripe formed by white dots, and its advertisement call; a long trill (up to almost 40 s) consisting of pairs of very short pulses. The discovery of these two apparently microendemic species suggests that additional Anomaloglossus species remain to be described in the Guiana Shield. Both species should be considered critically endangered given their seemingly reduced range size, association with highland habitat, and the anthropogenic pressure they currently face.

  18. Genetic variability of populations of Nyssomyia neivai in the Northern State of Paraná, Brazil

    Science.gov (United States)

    Gasparotto, Jaqueline de Carvalho; da Costa-Ribeiro, Magda Clara Vieira; Thomaz-Soccol, Vanete; Liebel, Sandra Mara Rodrigues da Silva; Neitzke-Abreu, Herintha Coeto; Reinhold-Castro, Kárin Rosi; Cristovão, Edilson Colhera; Teodoro, Ueslei

    2017-01-01

    ABSTRACT The genetic study of sandfly populations needs to be further explored given the importance of these insects for public health. Were sequenced the NDH4 mitochondrial gene from populations of Nyssomyia neivai from Doutor Camargo, Lobato, Japira, and Porto Rico, municipalities in the State of Paraná, Brazil, to understand the genetic structure and gene flow. Eighty specimens of Ny. Neivai were sequenced, 20 from each municipality, and 269 base pairs were obtained. A total of 27 haplotypes and 28 polymorphic sites were found, along with a haplotypic diversity of 0.80696 and a nucleotide diversity of 0.00567. Haplotype H5, with 33 specimens, was the most common among the four populations. Only haplotypes H5 and H7 were present in all four populations. The population from Doutor Camargo showed the highest genetic diversity, and only this population shared haplotypes with those from the other municipalities. The highest number of haplotypes was sheared with Lobato which also had the highest number of unique haplotypes. This probably occurred because of constant anthropic changes that happened in the environment during the first half of the twentieth century, mainly after 1998. There was no significant correlation between genetic and geographical distances regarding these populations. However, the highest genetic and geographical distances, and the lowest gene flow were observed between Japira and Porto Rico. Geographical distance is a possible barrier between these municipalities through the blocking of haplotype sharing. PMID:28380111

  19. Genetic variability of populations of Nyssomyia neivai in the Northern State of Paraná, Brazil

    Directory of Open Access Journals (Sweden)

    Jaqueline de Carvalho Gasparotto

    Full Text Available ABSTRACT The genetic study of sandfly populations needs to be further explored given the importance of these insects for public health. Were sequenced the NDH4 mitochondrial gene from populations of Nyssomyia neivai from Doutor Camargo, Lobato, Japira, and Porto Rico, municipalities in the State of Paraná, Brazil, to understand the genetic structure and gene flow. Eighty specimens of Ny. Neivai were sequenced, 20 from each municipality, and 269 base pairs were obtained. A total of 27 haplotypes and 28 polymorphic sites were found, along with a haplotypic diversity of 0.80696 and a nucleotide diversity of 0.00567. Haplotype H5, with 33 specimens, was the most common among the four populations. Only haplotypes H5 and H7 were present in all four populations. The population from Doutor Camargo showed the highest genetic diversity, and only this population shared haplotypes with those from the other municipalities. The highest number of haplotypes was sheared with Lobato which also had the highest number of unique haplotypes. This probably occurred because of constant anthropic changes that happened in the environment during the first half of the twentieth century, mainly after 1998. There was no significant correlation between genetic and geographical distances regarding these populations. However, the highest genetic and geographical distances, and the lowest gene flow were observed between Japira and Porto Rico. Geographical distance is a possible barrier between these municipalities through the blocking of haplotype sharing.

  20. Landscape structure in the northern coast of Paraná state, a hotspot for the brazilian Atlantic Forest conservation

    Directory of Open Access Journals (Sweden)

    Érico Emed Kauano

    2012-10-01

    Full Text Available The "Serra do Mar" region comprises the largest remnant of the Brazilian Atlantic Forest. The coast of the Paraná State is part of the core area of the "Serra do Mar" corridor and where actions for biodiversity conservation must be planned. In this study we aimed at characterizing the landscape structure in the APA-Guaraqueçaba, the largest protected area in this region, in order to assist environmental policies of this region. Based on a supervised classification of a mosaic of LANDSAT-5-TM satellite images (from March 2009, we developed a map (1:75,000 scale with seven classes of land use and land cover and analyzed the relative quantities of forests and modified areas in slopes and lowlands. The APA-Guaraqueçaba is comprised mainly by the Dense Ombrophilous Forest (68.6% of total area and secondary forests (9.1%, indicating a forested landscape matrix; anthropogenic and bare soil areas (0.8% and the Pasture/Grasslands class (4.2% were less representative. Slopes were less fragmented and more preserved (96.3% of Dense Ombrophilous Forest and secondary forest than lowlands (71.3%, suggesting that restoration initiatives in the lowlands must be stimulated in this region. We concluded that most of the region sustains well-conserved ecosystems, highlighting the importance of Paraná northern coast for the biodiversity maintenance of the Atlantic Forest.

  1. Effects of acacia senegal (L.,Willd.) on sandy soils: A case study of El damokeya forest, Northern Kordofan State

    International Nuclear Information System (INIS)

    Ahmed, D. M; Nimer, A. M.

    2002-01-01

    Soil properties were studied in El Damokeya forest, located at 30 km east of Elobeid town, Northern Kordofan State, during the rainy season of 1998. The aim was to characterize the soils of the area and to examine the effects of Acacia senegal plantations on the soils physical and chemical properties. The results showed that the soils were sandy, weakly structured, yellowish-red, neutral and poor in nutrient content, and that Acacia senegal plantations had induced considerable changes in the soil morphological, physical and chemical properties. The soil became more differentiated, with a third layer clearly discernible. No change had occurred in the soil texture. But, it became well structured with stable aggregates. Its organic matter content had been augmented to about one and half times, deeply incorporated and stained the whole profile with darker hues. The soil reaction became slightly acidic (ph 6.3). The exchange capacity was improved qualitatively and quantitatively. Thus, cation exchange capacity values increased from 2.8 in the bare land to 4.0 meq/100g soil under the forest, and the soil was saturated to 98% with base cations. The major nutrient elements (N,P, K, Ca, Mg, Fe) had generally increased with various proportions ranging from 10% to more than 130%, but only Ca showed significant difference at P=0.05. Among the trace elements, Cu and Co had significantly decreased in the forest soil, but Zn and Mn had increased to about 100%.(Author)

  2. Status of sheep sera to bluetongue, peste des petits ruminants and sheep pox in a few northern states of India

    Directory of Open Access Journals (Sweden)

    Vinayagamurthy Balamurugan

    2008-09-01

    Full Text Available Bluetongue (BT, peste des petits ruminants (PPR and sheep pox are the most economically important viral diseases of sheep in India. Serum samples obtained from sheep in five northern states of the country were screened for antibody against these agents to explore the extent of spread of these infections. A total of 516 serum samples were screened for the presence of antibodies against BT and PPR viruses. Of these, 155 samples were also tested for antibodies against sheep pox virus. BT antibodies were found in 293 (56.8% animals, PPR virus antibodies in 215 (41.7% and sheep pox virus antibodies in 106 (68.3%. Of the serum samples tested, 25.2% were positive for antibodies against all three viruses. These findings clearly demonstrated not only the enzootic nature of disease, but also the co-existence of antibodies to more than one of these viruses which would indicate that concurrent infections were common. Therefore, control measures should focus in combating all three diseases simultaneously by exploring the possibility of a trivalent vaccine or the use of multiple genes expressing vectored vaccine.

  3. Evaluation of normal values of Ca-15-3 and PSA for the people of Northern State of Sudan

    International Nuclear Information System (INIS)

    Bafaraj, S.M.I.

    2007-06-01

    In this study blood samples were collected from 665 female from Northern State of Sudan, 606 of those females are married and only 59 females are singles. These samples were used to estimate the normal CA 15-3 which was found to be similar to the international one and that is ranged from zero to 35 mlu/ml. When the same techniques were used using samples collected from patients who have been clinically diagnosed as breast cancer, CA 15-3 levels were found to be higher in most of these samples and the other finding is that the risk of getting breast cancer is starting from age 25 years. On the other hand , 650 males were participated in this study to estimate normal value of PSA, of them 553 were married. And again the levels estimated for this marker was found to be the same as that used globally which is ranged from zero to four ng/ml. When blood samples from prostate patients were assayed for PSA, the results showed high levels of this marker in almost all the samples as expected but the important finding is that prostate cancer is age depend. At the end of the project, many recommendations highlighted to be considered in the near future.(Author)

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

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

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

  5. The European Pressurized Water Reactor (EPR). State of the art after the preliminary design phase

    International Nuclear Information System (INIS)

    Bouteille, F.; Schneider, D.

    2002-01-01

    The European Pressurized Water Reactor (EPR) is an evolutionary development of the pressurized water reactor product lines built by Framatome and Siemens in France and Germany. Under the technical leadership of both nuclear power plant suppliers (now merged in Framatome ANP, a joint venture of AREVA and Siemens) the future-oriented plant concepts was developed in close cooperation with German and French utilities and in compliance with the European Utility Requirements. The EPR has safety features with which even extremely improbable, beyond design-basis events can be controlled and their effects can be limited to such an extent that no emergency response actions need be taken outside of the immediate plant site. This also means that safety systems prevent containment failure even in the improbable case of a core melt. This was confirmed by the French and German reactor safety authorities. The selected high thermal output also insures the economic viability of the innovative reactor concept, so that the power generation costs which can be achieved with the EPR will be absolutely competitive with those of fossil energy carriers. Framatome ANP has thus developed a pressurized water reactor ready for offer at the right time, which can completely fulfill the most rigorous requirements in terms of nuclear safety and economy. (Author)

  6. Safety-evaluation report related to the renewal of the operating license for the research reactor at the Iowa State University (Docket No. 50-116)

    International Nuclear Information System (INIS)

    1983-09-01

    This Safety Evaluation Report for the application filed by the Iowa State University (ISU) for a renewal of the Class 104 Operating License R-59 to continue to operate its Argonaut-type research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Iowa State University, and is located on the ISU campus in Ames, Story County, Iowa. The staff concludes that the reactor facility can continue to be operated by ISU without endangering the health and safety of the public. The principal matters reviewed are: design, testing, and performance of the reactor components and systems; the expected consequences of credible accidents; the licensee's management organization; the method used for the control of radiological effluents; the licensee's technical specifications; financial data and information; the physical protection program; procedures for training reactor operators; and emergency plans. 11 references, 15 figures, 13 tables

  7. State of development of gas cooled reactors in the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Mosevitskij, I.S.

    1991-01-01

    In the context of the programme for the development of gas-cooled reactors in the USSR it is reported that pilot plants with VGR-50 MW(el) and VG-400 MW(el) have been developed up to the stage of engineering design and that now the efforts are concentrated on the project of pilot-commercial reactor plant VGM (PCRP VGM) of a modular type with unit thermal power of 200-250 MW. The installation is designed to solve the main scientific and engineering problems of construction of high-temperature gas-cooled reactors, to test equipment components, and to show advantages of the given type of installations having the enhanced safety and capability to generate high-potential heat. The status of work on the PCRP VGM project is described. 3 refs, 1 fig., 1 tab

  8. Perturbative methods for sensitivity calculation in safety problems of nuclear reactors: state-of-the-art

    International Nuclear Information System (INIS)

    Lima, Fernando R.A.; Lira, Carlos A.B.O.; Gandini, Augusto

    1995-01-01

    During the last two decades perturbative methods became an efficient tool to perform sensitivity analysis in nuclear reactor safety problems. In this paper, a comparative study taking into account perturbation formalisms (Diferential and Matricial Mthods and generalized Perturbation Theory - GPT) is considered. Then a few number of applications are described to analyze the sensitivity of some functions relavant to thermal hydraulics designs or safety analysis of nuclear reactor cores and steam generators. The behaviours of the nuclear reactor cores and steam generators are simulated, respectively, by the COBRA-IV-I and GEVAP codes. Results of sensitivity calculations have shown a good agreement when compared to those obtained directly by using the mentioned codes. So, a significative computational time safe can be obtained with perturbative methods performing sensitivity analysis in nuclear power plants. (author). 25 refs., 5 tabs

  9. Methodology for development of health physics procedures at research reactors in agreement states

    International Nuclear Information System (INIS)

    Woodard, R.C.; Bauer, T.L.; Wehring, B.W.

    1991-01-01

    The University of Texas at Austin is awaiting final license approval to operate a new 1 MW TRIGA reactor for teaching and research. All reactor and laboratory operations, experiments, and monitoring are carried out under health physics procedures that address to ensure consideration of all applicable documents as references in order to comply with the regulations and accepted good practices. This paper examines the development of one procedure Radioactive Material Control by use of the method. The process is examined as a tool to apply to any health physics procedure development. Further discussion focuses on the regulatory anomalies observed during development of the procedure and presents the arguments for the authors resolution of these issues. The design of the reactor facility is also detailed to allow for understanding of the problems encountered during procedural development

  10. Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

    International Nuclear Information System (INIS)

    Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

    2006-01-01

    Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material 233 U which shows superior characteristics in the thermal region. The Shippingport reactor used 233 U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, η value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required 233 U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of 233 U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

  11. Light water reactor (LWR) innovation needs in the United States: The Massachusetts Institute of Technology LWR innovation project

    International Nuclear Information System (INIS)

    Golay, M.W.

    1988-01-01

    A major effort under way within the Massachusetts Institute of Technology (MIT) Engineering School is focused on the contributions that technology innovation can make in revitalizing nuclear power in the United States. A principal component of this effort is a project to improve the designs of the next generation of light water reactors (LWRs) with emphasis on achieving improved capacity factors and safety, and reducing the construction duration. The motivation for this overall effort is to prevent the nuclear option from being unnecessarily lost by being available only in uneconomic configurations. In considering how to advance this effort, the authors focused on refining the designs of new reactors because this is the area where the greatest opportunities for improvements exist

  12. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles

  13. United States Department of Energy breeder reactor staff training domestic program

    International Nuclear Information System (INIS)

    1984-01-01

    Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described

  14. Conceptual design study of quasi-steady state fusion experimental reactor (FER-Q), part 2

    International Nuclear Information System (INIS)

    1985-12-01

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 FER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included: heating/current drive system, plasma position control, power supply, diagnostics, neutronics, blanket test module, repair and maintenance and safety. (author)

  15. The present state of development and the future of the high-temperature reactor in the United States of America

    International Nuclear Information System (INIS)

    Simon, W.A.; Chi, H.W.

    1982-01-01

    The American prototype high-temperature reactor at Fort St. Vrain has been operating successfully for years. To date it has produced more than 3.000.000.000 kilowatt hours of electricity and a short while ago was cleared for operation at full load. Operating experience justifies expectations that the combined cycle HTR plant of 2240 MW thermal output favoured by the US Government and industry will offer significant economic advantages. (orig.) [de

  16. Safety evaluation report related to the renewal of the operating license for the Washington State University TRIGA reactor. Docket No. 50-27

    International Nuclear Information System (INIS)

    1982-05-01

    This Safety Evaluation Report for the application filed by the Washington State University (WSU) for a renewal of operating license number R-76 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Washington State University and is located on the WSU campus in Pullman, Whitman County, Washington. The staff concludes that the TRIGA reactor facility can continue to be operated by WSU without endangering the health and safety of the public

  17. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  18. Welding of sule elements for nuclear reactors with solid state YAG laser using instrumentated testing equipments

    International Nuclear Information System (INIS)

    Bourgault, F.; Lacoste, J.; Schley, R.; Kluzinski, C.; Piednoir, P.

    1985-09-01

    The instrumentation of the equipment for carrying out safety tests on fuel elements for nuclear reactors requires special thermocouples adapted to the prevailing agressive medium. The investigations described deal essentially with the operational and metallurgical weldability tests out on the safety test zircaloy piping in the pressurized water circuit (PHEBUS-programme) [fr

  19. State of the art on reactor designs for solar gasification of carbonaceous feedstock

    DEFF Research Database (Denmark)

    Puig Arnavat, Maria; Tora, E.A.; Bruno, J.C.

    2013-01-01

    to produce high quality synthesis gas with a higher output per unit of feedstock and that allows for the chemical storage of solar energy in the form of a readily transportable fuel, among other advantages. The present paper describes the latest advances in solar thermochemical reactors for gasification...

  20. Calculation of the Thermal State of the Graphite Moderator of the RBMK Reactor

    Directory of Open Access Journals (Sweden)

    Vorobiev Alexander V.

    2017-01-01

    Full Text Available This work is devoted to study the temperature field of the graphite stack of the RBMK reactor. In work was analyzed the influence of contact pressure between the components of the masonry on the temperature of the graphite moderator.

  1. 25 years Rossendorf research reactor RFR - retrospect, state of the art and plans

    International Nuclear Information System (INIS)

    Hieronymus, W.

    1982-01-01

    Starting from a review of the essential stages of the RFR history, future tasks are presented primarily from the point of view of operating the reactor past the year 1987. It is reported on the achieved level of utilization and on the organization of plant management. (author)

  2. MEaSUREs Northern Hemisphere State of Cryosphere Daily 25km EASE-Grid 2.0

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set reports the location of Northern Hemisphere snow cover and sea ice extent, the status of melt onset across Greenland and Artic sea ice, and the level...

  3. Gender difference in early initiation of methamphetamine use among current methamphetamine users in Muse, Northern Shan State, Myanmar.

    Science.gov (United States)

    Saw, Yu Mon; Saw, Thu Nandar; Yasuoka, Junko; Chan, Nyein; Kham, Nang Pann Ei; Khine, Wint; Cho, Su Myat; Jimba, Masamine

    2017-05-08

    Globally, methamphetamine (MA) use is a significant public health concern due to unprecedented health effects of its use. However, gender similarities and differences in early age of MA initiation and its risk factors among current MA users have been understudied in a developing country setting. A community-based, cross-sectional study was conducted using a computer assisted self-interviewing program from January to March 2013 in Muse, Northern Shan State, Myanmar. A total of 1362 (775 male and 587 female) self-reported current MA users aged between 18 and 35 years were recruited using respondent-driven sampling. Two gender-stratified multiple logistic regression models (models I and II) were done for analysis. For similarities, 73.0% of males and 60.5% of females initiated MA before their 18th birthday. The early age of MA initiation was positively associated with the reasons and places of the first time MA use among both genders. For differences, males [hazard ratio 1.35; 95% confidence interval, 1.18-1.54] had a significantly higher risk than females to initiate MA at earlier age. Among male users, participants who had bisexual/homosexual preferences were more likely to initiate MA use earlier. In contrast, female users who exchanged sex for money and/or drugs were more likely to initiate MA in earlier age. More than 60.0% of male and female participants initiated MA use early; however, males initiated use earlier than females. Although similarities were found among both genders, differences found in key risk factors for early age MA initiation suggest that gender-specific, MA prevention programs are urgently needed in Myanmar.

  4. Steady State Simulation of Two-Gas Phase Fluidized Bed Reactors in Series for Producing Linear Low Density Polyethylene

    Directory of Open Access Journals (Sweden)

    Ali Farhangiyan Kashani

    2012-12-01

    Full Text Available A linear low density polyethylene (LLDPE production process, including two- fuidized bed reactors in series (FBRS and other process equipment, was completely simulated by Aspen Polymer Plus software. Fluidized bed reactors were considered as continuous stirred tank reactors (CSTR consisted of polymer and gas phases. POLY-SRK and NRTL-RK equations of state were used to describe polymer and non-polymer streams, respectively. In this simulation, a kinetic model, based on a double active site heterogeneous Ziegler-Natta catalyst was used for simulation of LLDPE process consisting of two FBRS. Simulator using this model has the capability to  predict a number of  principal characteristics of LLDPE such as melt fow index (MFI, density, polydispersity index, numerical and weight average molecular weights (Mn,Mw and copolymer molar fraction (SFRAC. The results of the simulation were compared with industrial plant data and a good agreement was observed between the predicted model and plant data. The simulation results show the relative error of about 0.59% for prediction of polymer mass fow and 2.67% and 0.04% for prediction of product MFI and density, respectively.

  5. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

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

  6. The design, safety and project development status of the modular high temperature gas-cooled reactor in the United States

    International Nuclear Information System (INIS)

    Mears, L.D.; Dean, R.A.

    1987-01-01

    The cooperative government and industry Modular High Temperature Gas-Cooled Reactor (MHTGR) Program in the United States has advanced a 350 MW(t) plant design through the conceptual development stage. The system incorporates an annular core of prismatic fuel elements within a steel pressure vessel connected, in a side-by-side arrangement, by a concentric duct to a second steel vessel containing a steam generator and helium coolant circulator. The reference plant design consists of four reactor modules installed in separate below-grade silos, providing steam to two conventional turbine generators. The nominal net plant output is 540 MW(e). The small reactor system takes unique advantage of the high temperature capability of the refractory coated fuel and the large thermal inertia of the graphite moderator to provide a design capable of withstanding a complete loss of active core cooling without causing excessive core heatup and significant release of fission products from the fuel. Present program activities are concentrated on interactions with the Nuclear Regulatory Commission aimed at obtaining a Licensability Statement. A project initiative to build a prototype plant which would demonstrate the MHTGR-unique licensing process, plant performance, costs and schedule plus establish an industrial infrastructure to proceed with follow-on commercial MHTGR plants by the turn of the century, is being undertaken by the utility/vendor participants (author)

  7. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    International Nuclear Information System (INIS)

    1997-04-01

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

  8. Progress of the United States foreign research reactor spent nuclear fuel acceptance program. Reduced enrichment for research and test reactors conference 2002

    International Nuclear Information System (INIS)

    Clapper, Maureen

    2002-01-01

    Foreign Research Reactor Spent nuclear fuel Acceptance Program is actively working with research reactors to accept eligible material before the Acceptance Policy proper expires in 2006. Reactors/governments wishing to participate should contact US immediately if they have not done so already. Program operations are changing to adapt to new challenges. We continue to promote the importance of this Program to senior management in the Department of Energy

  9. Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

    International Nuclear Information System (INIS)

    Gou Junli; Qiu Suizheng; Su Guanghui; Jia Dounan

    2006-01-01

    This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation. (authors)

  10. Dynamical and technological consequences of multiple isolas of steady states in a catalytic fluidised-bed reactor

    Directory of Open Access Journals (Sweden)

    Bizon Katarzyna

    2017-09-01

    Full Text Available Steady-state characteristics of a catalytic fluidised bed reactor and its dynamical consequences are analyzed. The occurrence of an untypical steady-state structure manifesting in a form of multiple isolas is described. A two-phase bubbling bed model is used for a quantitative description of the bed of catalyst. The influence of heat exchange intensity and a fluidisation ratio onto the generation of isolated solution branches is presented for two kinetic schemes. Dynamical consequences of the coexistence of such untypical branches of steady states are presented. The impact of linear growth of the fluidisation ratio and step change of the cooling medium temperature onto the desired product yield is analyzed. The results presented in this study confirm that the identification of a region of the occurrence of multiple isolas is important due to their strong impact both on the process start-up and its control.

  11. Calibration of A Prompt Gamma Neutron Activation Analysis (PGNAA) Facility: Experience at the Oregon State University TRIGA Reactor

    International Nuclear Information System (INIS)

    Norlida Yussup

    2011-01-01

    A prompt gamma neutron activation analysis (PGNAA) facility at Oregon State University (OSU) TRIGA reactor has been built in year 2008 and been operated since then. PGNAA is a technique used to determine the presence and quantity of trace elements such as boron, hydrogen and carbon which are more difficult to detect with other neutron analysis method. A calibration is essential to ensure the system works as required and the output is valid and reliable. The calibration was carried out by using Standard Reference Material (SRM). Besides, background data was also acquired for comparisons and analysis. The results are analyzed and discussed in this paper. (author)

  12. Gas reactor international cooperative program interim report: United States/Federal Republic of Germany nuclear licensing comparison

    International Nuclear Information System (INIS)

    1978-09-01

    In order to compare US and FRG Nuclear Licensing, a summary description of United States Nuclear Licensing is provided as a basis. This is followed by detailed information on the participants in the Nuclear Licensing process in the Federal Republic of Germany (FRG). FRG licensing procedures are described and the rules and regulations imposed are summarized. The status of gas reactor licensing in both the U.S. and the FRG is outlined and overall conclusions are drawn as to the major licensing differences. An appendix describes the most important technical differences between US and FRG criteria

  13. Cellulase production by Trichoderma harzianum in static and mixed solid-state fermentation reactors under nonaseptic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Deschamps, F.; Giuliano, C.; Asther, M.; Huet, M.C.; Roussos, S.

    1985-09-01

    Cellulase production from lignocellulosic materials was studied in solid-state cultivation by both static and mixed techniques under nonaseptic conditions. The effects of fermentation conditions, such as moisture content, pH, temperature, and aeration, on cellulase production by Trichoderma harzianum using a mixture of wheat straw (80%) and bran (20%) were investigated. With a moisture content of 74% and a pH of 5.8, 18 IU filter paper activity and 198 IU endoglucanase activity/g initial substrate content were obtained in 66 hours. The extension from static column cultivation to stirred tank reactor of 65 l capacity gave similar yields of cellulase.

  14. State of Art Report for the OECD-NEA Loss-of-Forced Cooling (LOFC) Test Project using HTTR Reactor

    International Nuclear Information System (INIS)

    Jun, Ji Su

    2011-05-01

    The OECD/NEA Project is planned to perform the LOFC (Loss Of Forced Cooling) test using the HTTR (High Temperature engineering Test Reactor) in Japan from 31 March 2011 to 31 March 2013 in order to obtain the data for the code validation of the VHTR safety analysis. Based on the Project Agreement Document, this report gives a description of the HTTR-LOFC test, HTTR test facility, project schedule and deliverable items as the technical state art of the project, and appends the full translation of the project agreement articles on the project management

  15. Participation in the United States Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  16. Sodium-cooled reactors, objectives, achieved technical state and development trends

    International Nuclear Information System (INIS)

    Wolff, U.

    1988-01-01

    The use of fossil fuels to cover the future world-wide energy demand alone would rapidly deplete these ressources, especially oil and gas. Today's knowledge suggests the enhanced exploitation of solar energy, nuclear fusion and the application of uranium in sodium-cooled breeder reactors as the alternative energies offering a great potential. The sodium-cooled reactor outdistances the other options in terms of development. Its technical feasibility and safe operation have been verified and its profitability appears to be possible when using today's technology. The verification of its profitability while maintaining a high safety level is the overriding task for the future. The paper discusses corresponding activities in the USA, the USSR, Japan and Western Europe. (orig.) [de

  17. State of the art seismic analysis for CANDU reactor structure components using condensation method

    Energy Technology Data Exchange (ETDEWEB)

    Soliman, S A; Ibraham, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The reactor structure assembly seismic analysis is a relatively complex process because of the intricate geometry with many different discontinuities, and due to the hydraulic attached mass which follows the structure during its vibration. In order to simulate reasonably accurate behaviour of the reactor structure assembly, detailed finite element models are generated and used for both modal and stress analysis. Guyan reduction condensation method was used in the analysis. The attached mass, which includes the fluid mass contained in the components plus the added mass which accounts for the inertia of the surrounding fluid entrained by the accelerating structure immersed in the fluid, was calculated and attached to the vibrating structures. The masses of the attached components, supported partly or totally by the assembly which includes piping, reactivity control units, end fittings, etc. are also considered in the analysis. (author). 4 refs., 6 tabs., 4 figs.

  18. Survey of considerations involved in introducing CANDU reactors into the United States

    International Nuclear Information System (INIS)

    Till, C.E.; Bohn, E.M.; Chang, Y.I.; van Erp, J.B.

    1977-01-01

    The important issues that must be considered in a decision to utilize CANDU reactors in the U.S. are identified in this report. Economic considerations, including both power costs and fuel utilization, are discussed for the near and longer term. Safety and licensing considerations are reviewed for CANDU-PHW reactors in general. The important issues, now and in the future, associated with power generation costs are the capital costs of CANDUs and the factors that impact capital cost comparisons. Fuel utilization advantages for the CANDU depend upon assumptions regarding fuel recycle at present, but the primary issue in the longer term is the utilization of the thorium cycle in the CANDU. Certain safety features of the CANDU are identified as intrinsic to the concept and these features must be examined more fully regarding licensability in the U.S

  19. Participation in the United States Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993

  20. Experimental and theoretical comparison of fuel temperature and bulk coolant characteristics in the Oregon State TRIGA reactor during steady state operation

    Energy Technology Data Exchange (ETDEWEB)

    Marcum, W.R., E-mail: marcumw@engr.orst.ed [Oregon State University, Department of Nuclear Engineering and Radiation Health Physics, 116 Radiation Center, Corvallis, OR 97330 (United States); Woods, B.G.; Reese, S.R. [Oregon State University, Department of Nuclear Engineering and Radiation Health Physics, 116 Radiation Center, Corvallis, OR 97330 (United States)

    2010-01-15

    In September of 2008 Oregon State University (OSU) completed its core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Experimental bulk coolant temperatures were collected in various locations throughout the Oregon State TRIGA Reactor (OSTR) core in order to supplement the validity of the numerical thermal hydraulic results produced in RELAP5-3D Version 2.4.2. Axial bulk coolant temperature distributions were collected by acquiring discrete thermocouple measurements in individual subchannel locations during steady state operation at 1.0 MW{sub th}. The experimental axial temperature distribution collected was compared to one-channel, two-channel, and eight-channel RELAP5-3D models and found to match within 11.94%, 11.69%, and 8.78%, respectively, on average. Comparisons to similar studies were made based on a dimensional analysis of fluid body forces in the discrete core locations, indicating that the chosen approach produces conservative results for use in the OSTR safety analysis.

  1. Core-adjacent instrumentation systems for pebble bed reactors for process heat application - state of planning

    International Nuclear Information System (INIS)

    Benninghofen, G.; Serafin, N.; Spillekothen, H.G.; Hecker, R.; Brixy, H.; Serpekian, T.

    1982-06-01

    Planning and theoretical/experimental development work for core surveillance instrumentation systems is being performed to meet requirements of pebble bed reactors for process heat application. Detailed and proved instrumentation concepts are now available for the core-adjacent instrumentation systems. The current work and the results of neutron flux measurements at high temperatures are described. Operation devices for long-term accurate gas outlet temperature measurements up to approximately 1423 deg. K will also be discussed. (author)

  2. Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Buiron, Laurent

    2015-01-01

    In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

  3. Extent of Parent-Teacher Association Involvement in the Implementation of Universal Basic Education Program in Primary Schools in Northern Senatorial District of Ondo State, Nigeria

    Directory of Open Access Journals (Sweden)

    Chidi Nnebedum

    2018-05-01

    Full Text Available Pupil’ absenteeism and lateness to school, dilapidated and shortage of relevant facilities in primary schools in the Northern Senatorial District of Ondo State seems to suggest lapses in parent-teacher association involvement in school affairs. This prompted the researchers to ascertain the extent of parent-teacher association (PTA involvement in the implementation of universal basic education program in primary schools in the Northern Senatorial District of Ondo State. Three research questions guided the study and three null hypotheses were tested. The descriptive survey research design was adopted for the study. The population of the study was comprised of all 250 head teachers and all PTA members at all 250 primary schools in the Northern Senatorial District. Multiple stage sampling technique was used to sample 205 respondents made up of 75 head teachers and 130 PTA members. The researchers developed an instrument titled “Parent-Teacher Association Involvement in School Questionnaire (PTAISQ” which was used for data collection. The instrument was validated by three experts. The reliability of the instrument was ascertained using Cronbach alpha and it yielded an overall reliability coefficient value of .76. Mean and standard deviation were used to answer the research questions, while t-test was used to test the null hypotheses at .05 level of significance. The findings of the study revealed among others that the extent of PTA involvement in maintenance of facilities in the implementation of universal basic education program in primary schools in the Northern Senatorial District of Ondo State was high. Based on the findings, recommendations were made and conclusions were drawn.

  4. Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

  5. Design constraints for rf-driven steady-state tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1979-02-01

    Plasma current density profiles are computed due to electron Landau damping of lower hybrid waves launched into model tokamak density and temperature profiles. The total current and current profile shape are chosen consistent with magnetohydrodynamic equilibrium for a variety of temperature and density distributions and plasma beta values. Surface current equilibria appear attractive and are accessible to waves with n/sub z/ as low as 1.2. By suitably choosing the spectrum location and width it is possible to drive the 9.8 MA current of a 7.0-m reactor with as little as 2.8% of the fusion power recirculated as rf input from the waveguides

  6. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  7. Mid-21st-century climate changes increase predicted fire occurrence and fire season length, Northern Rocky Mountains, United States

    Science.gov (United States)

    Riley, Karin L.; Loehman, Rachel A.

    2016-01-01

    Climate changes are expected to increase fire frequency, fire season length, and cumulative area burned in the western United States. We focus on the potential impact of mid-21st-century climate changes on annual burn probability, fire season length, and large fire characteristics including number and size for a study area in the Northern Rocky Mountains. Although large fires are rare they account for most of the area burned in western North America, burn under extreme weather conditions, and exhibit behaviors that preclude methods of direct control. Allocation of resources, development of management plans, and assessment of fire effects on ecosystems all require an understanding of when and where fires are likely to burn, particularly under altered climate regimes that may increase large fire occurrence. We used the large fire simulation model FSim to model ignition, growth, and containment of wildfires under two climate scenarios: contemporary (based on instrumental weather) and mid-century (based on an ensemble average of global climate models driven by the A1B SRES emissions scenario). Modeled changes in fire patterns include increased annual burn probability, particularly in areas of the study region with relatively short contemporary fire return intervals; increased individual fire size and annual area burned; and fewer years without large fires. High fire danger days, represented by threshold values of Energy Release Component (ERC), are projected to increase in number, especially in spring and fall, lengthening the climatic fire season. For fire managers, ERC is an indicator of fire intensity potential and fire economics, with higher ERC thresholds often associated with larger, more expensive fires. Longer periods of elevated ERC may significantly increase the cost and complexity of fire management activities, requiring new strategies to maintain desired ecological conditions and limit fire risk. Increased fire activity (within the historical range of

  8. Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pomorski, Michal; Mer-Calfati, Christine [CEA-LIST, Diamond Sensors Laboratory, 91191, Gif-sur-Yvette (France); Foulon, Francois [CEA, National Institute for Nuclear Science and Technology, 91191, Gif-sur-Yvette (France); Sklenka, Lubomir; Rataj, Jan; Bily, Tomas [Department of Nuclear Reactors,Faculty of Nuclear Science and Physical Engineering, Czech Technical University, V. Holesovickach 2, 180 00 PRAHA 8 (Czech Republic)

    2015-07-01

    Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in

  9. Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

    International Nuclear Information System (INIS)

    Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois; Sklenka, Lubomir; Rataj, Jan; Bily, Tomas

    2015-01-01

    Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm 2 . Detectors with surfaces up to 1 cm 2 can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm 2 , with the possibility to enlarge the surface of the detector up to 1 cm 2 . These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The

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

    Science.gov (United States)

    Kadyk, Thomas; Eikerling, Michael

    2015-08-14

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

  11. An assessment of the downturn in the forest products sector in the northern region of the United States

    Science.gov (United States)

    C.W. Woodall; W.G. Luppold; P.J. Ince; R.J. Piva; K.E. Skog

    2012-01-01

    The forest industry within the northern region of the U.S. has declined notably in employment, mill numbers, wood consumption, and forest harvests since 2000…a downturn exacerbated by the recession of 2007 to 2009. Longer term industrial decline (since 2000) has been evidenced by reductions in secondary products (e.g., furniture) and print paper manufacturing which can...

  12. Solid State Track Recorder fission rate measurements in low power light water reactor pressure vessel mockups

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Kellogg, L.S.

    1985-01-01

    The results of extensive SSTR measurements made at the Pool Critical Assembly (PCA) facility at Oak Ridge National Laboratory have been reported previously. Measurements were made at key locations in PCA which is an idealized mockup of the water gap, thermal shield, pressure vessel geometry of a light water reactor. Recently, additional SSTR fission rate measurements have been carried out for 237-Np, 238-U, and 235-U in key locations in the NESTOR Shielding and Dosimetry Improvement Program (NESDIP) mockup facility located at Winfrith, England. NESDIP is a replica of the PCA facility, and comparisons will be made between PCA and NESDIP measurements. The results of measurements made at the engineering mockup at the VENUS critical assembly at CEN/SCK, Mol, Belgium will also be reported. Measurements were made at selected radial and azimuthal locations in VENUS, which models the in-core and near-core regions of a pressurized water reactor. Comparisons of absolute SSTR fission rates with absolute fission rates made with the Mol miniature fission chamber will be reported. Absolute fission rate comparisons have also been made between the NBS fission chamber, radiometric fission foils, and SSTRs, and these results will be summarized

  13. The design status of the United States Department of Energy modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Mills, Raymond R. Jr.

    1990-01-01

    The U.S. Department of Energy's Modular High Temperature Gas Cooled Reactor (MHTGR) is being designed using a systems engineering approach referred to as the integrated approach. The top level requirement for the plant is that it provides safe, reliable, economical energy. The safety requirements are established by the U.S. Licensing Authorities, principally the Nuclear Regulatory Commission. The reliability and economic requirements associated with the top level functions have been established in close coordination and cooperation with the electrical utilities and other potential users, and the nuclear supply industry. The integrated approach uses functional analysis to define the functions and sub-functions for the plant and to identify quantitatively how the various functions must be fulfilled. The top four functions associated with the MHTGR are: maintain safe plant operation; maintain plant protection; maintain control of radionuclide release; maintain emergency preparedness. In addition to meeting all U.S. Regulatory Requirements this advanced reactor concept is being designed to meet the following requirements: do not require sheltering or evacuating of anyone outside the plant boundary of 425 meters as a result of normal or abnormal plant operation; do not require operator action in order to accomplish the above sheltering and evacuation objectives and the design must be insensitive to operator errors; utilize inherent characteristics of materials to develop passive safety features; provide very long times for corrective actions following the initiation of an abnormal event before plant damage would be incurred

  14. The current state of the Russian reduced enrichment research reactors program

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A. [and others

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

  15. A Singular Perturbation Problem for Steady State Conversion of Methane Oxidation in Reverse Flow Reactor

    Directory of Open Access Journals (Sweden)

    Aang Nuryaman

    2012-11-01

    Full Text Available The governing equations describing the methane oxidation process in reverse flow reactor are given by a set of convective-diffusion equations with a nonlinear reaction term, where temperature and methane conversion are dependent variables. In this study, the process is assumed to be one-dimensional pseudo homogeneous model and takes place with a certain reaction rate in which the whole process of reactor is still workable. Thus, the reaction rate can proceed at a fixed temperature. Under this condition, we restrict ourselves to solve the equations for the conversion only. From the available data, it turns out that the ratio of the diffusion term to the reaction term is small. Hence, this ratio is considered as small parameter in our model and this leads to a singular perturbation problem. In the vicinity of small parameter in front of higher order term, the numerical difficulties will be found. Here, we present an analytical solution by means of matched asymptotic expansions. Result shows that, up to and including the first order of approximation, the solution is in agreement with the exact and numerical solutions of the boundary value problem.

  16. Hydrochemical Regions of the Glacial Aquifer System, Northern United States, and Their Environmental and Water-Quality Characteristics

    Science.gov (United States)

    Arnold, Terri L.; Warner, Kelly L.; Groschen, George E.; Caldwell, James P.; Kalkhoff, Stephen J.

    2008-01-01

    The glacial aquifer system in the United States is a large (953,000 square miles) regional aquifer system of heterogeneous composition. As described in this report, the glacial aquifer system includes all unconsolidated geologic material above bedrock that lies on or north of the line of maximum glacial advance within the United States. Examining ground-water quality on a regional scale indicates that variations in the concentrations of major and minor ions and some trace elements most likely are the result of natural variations in the geologic and physical environment. Study of the glacial aquifer system was designed around a regional framework based on the assumption that two primary characteristics of the aquifer system can affect water quality: intrinsic susceptibility (hydraulic properties) and vulnerability (geochemical properties). The hydrochemical regions described in this report were developed to identify and explain regional spatial variations in ground-water quality in the glacial aquifer system within the hypothetical framework context. Data analyzed for this study were collected from 1991 to 2003 at 1,716 wells open to the glacial aquifer system. Cluster analysis was used to group wells with similar ground-water concentrations of calcium, chloride, fluoride, magnesium, potassium, sodium, sulfate, and bicarbonate into five unique groups. Maximum Likelihood Classification was used to make the extrapolation from clustered groups of wells, defined by points, to areas of similar water quality (hydrochemical regions) defined in a geospatial model. Spatial data that represented average annual precipitation, average annual temperature, land use, land-surface slope, vertical soil permeability, average soil clay content, texture of surficial deposits, type of surficial deposit, and potential for ground-water recharge were used in the Maximum Likelihood Classification to classify the areas so the characteristics of the hydrochemical regions would resemble the

  17. Functional role of fouling community on an artificial reef at the northern coast of Rio de Janeiro State, Brazil

    Directory of Open Access Journals (Sweden)

    Werther Krohling

    2006-12-01

    Full Text Available The northern coast of Rio de Janeiro State lacks natural consolidate substrates, making it a proper environment to the development of researches using artificial structures. After studies about the type of substrate, concrete seems to be the most appropriate for studying fouling community development. This research was carried out to investigate the functional role of biofouling in the development of the ichthyic community in the north of Rio de Janeiro State. Percentage data of the epibenthic organisms' coverage and samples of the fish community with gillnet and visual census showed that biofouling in artificial reefs might have more than one functional role, acting as a facilitator in the recruitment of fish species and as a link in the trophic marine chain. Through the increase of localized structural complexity provided by the reef itself and by the fouling organisms which act as "engineering species", additional protection options are offered to the ichthyic community, especially recruits. Also, the epibiont biomass represents an important link in the food web, acting either as a direct source or in the transference of energy to higher trophic levels. Through the relationship between the ichthyic and fouling communities we concluded that the functional role of the latter in artificial reef habitats could be characterized mainly as shelter and feeding grounds for few fish species.O litoral norte do Estado do Rio de Janeiro possui uma escassez de substrato consolidado natural tornando o ambiente propício para o desenvolvimento de pesquisas com estruturas artificiais. Após estudos conclusivos sobre o tipo de substrato, o concreto parece ser o mais apropriado para o desenvolvimento da comunidade incrustante. Novas pesquisas foram realizadas para investigar o papel funcional da bioincrustação no desenvolvimento da comunidade íctica no norte do Estado do Rio de Janeiro. Dados de porcentagem de cobertura dos organismos epibênticos e

  18. Comparison of the microbial communities in solid-state anaerobic digestion (SS-AD) reactors operated at mesophilic and thermophilic temperatures.

    Science.gov (United States)

    Li, Yueh-Fen; Nelson, Michael C; Chen, Po-Hsu; Graf, Joerg; Li, Yebo; Yu, Zhongtang

    2015-01-01

    The microbiomes involved in liquid anaerobic digestion process have been investigated extensively, but the microbiomes underpinning solid-state anaerobic digestion (SS-AD) are poorly understood. In this study, microbiome composition and temporal succession in batch SS-AD reactors, operated at mesophilic or thermophilic temperatures, were investigated using Illumina sequencing of 16S rRNA gene amplicons. A greater microbial richness and evenness were found in the mesophilic than in the thermophilic SS-AD reactors. Firmicutes accounted for 60 and 82 % of the total Bacteria in the mesophilic and in the thermophilic SS-AD reactors, respectively. The genus Methanothermobacter dominated the Archaea in the thermophilic SS-AD reactors, while Methanoculleus predominated in the mesophilic SS-AD reactors. Interestingly, the data suggest syntrophic acetate oxidation coupled with hydrogenotrophic methanogenesis as an important pathway for biogas production during the thermophilic SS-AD. Canonical correspondence analysis (CCA) showed that temperature was the most influential factor in shaping the microbiomes in the SS-AD reactors. Thermotogae showed strong positive correlation with operation temperature, while Fibrobacteres, Lentisphaerae, Spirochaetes, and Tenericutes were positively correlated with daily biogas yield. This study provided new insight into the microbiome that drives SS-AD process, and the findings may help advance understanding of the microbiome in SS-AD reactors and the design and operation of SS-AD systems.

  19. Best-estimate reactor core monitor using state feedback strategies to resolve uncertainties

    International Nuclear Information System (INIS)

    Martin, R.P.

    1997-01-01

    The development and demonstration of a new algorithm for quantifying uncertainty in best-estimate simulation codes has been investigated. Demonstration is given by way of a prototype reactor core monitor. The architecture of this monitor integrates a distributed parameter estimation technique and the infrastructure required to support this control theory-based algorithm into a production-grade best-estimate simulation code. The Kalman filter with the sequential least-squares parameter estimation algorithm has been extended for application into the computational environment of a best-estimate simulation code, i.e., RELAP5/DOE. In control system terminology this configuration can be thought of as a best-estimate observer

  20. United States Department of Energy projects related to reactor pressure vessel annealing optimization

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Nakos, J.T.

    1993-01-01

    Light water reactor pressure vessel (RPV) material properties reduced by long-term exposure to neutron irradiation can be recovered through a thermal annealing treatment. This technique to extend RPV life, discussed in this report, provides a complementary approach to analytical methodologies to evaluate RPV integrity. RPV annealing has been successfully demonstrated in the former Soviet Union and on a limited basis by the US (military applications only). The process of demonstrating the technical feasibility of annealing commercial US RPVs is being pursued through a cooperative effort between the nuclear industry and the US Department of Energy (USDOE) Plant Lifetime Improvement (PLIM) Program. Presently, two projects are under way through the USDOE PLIM Program to demonstrate the technical feasibility of annealing commercial US RPVS, (1) annealing re-embrittlement data base development and (2) heat transfer boundary condition experiments

  1. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  2. Use of the modular modeling system in the design of the Penn State Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    Smith, K.A.

    1988-12-01

    This study involves the design and subsequent transient analysis of the Penn State Advanced Light Water Reactor (PSU ALWR). The performance of the PSU ALWR is evaluated during small step changes in power and a turbine trip from full power without scram. The Modular Modeling System (MMS), developed by Babcock and Wilcox under a contract from the Electric Power Research Institute (EPRI), is a computer code designed for the simulation of nuclear and fossil power plants. MMS uses preprogrammed modules to represent specific power plant components such as pipes, pumps, steam generators, and a nuclear reactor. These components can then be connected in any manner the user desires providing certain simple interconnection rules are followed. In this study, MMS is used to develop computer models of both the PSU ALWR and a conventional PWR operating at the same power level. These models are then subjected to the transients mentioned above to evaluate the ability of the letdown-injection system to maintain primary system pressure. The transient response of the PSU ALWR and conventional PWR MMS models were compared to each other and whenever possible to actual plant transient data. 14 refs., 29 figs., 5 tabs

  3. Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

    Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

  4. 75 FR 23820 - Notice of Docketing of Amendment Request for Materials License No. SNM-2506; Northern States...

    Science.gov (United States)

    2010-05-04

    ... INFORMATION CONTACT: Pamela Longmire, Ph.D., Project Manager, Licensing Branch, Division of Spent Fuel Storage... Generating Plant (PINGP), Unit Nos. 1 and 2, site in Goodhue County, Minnesota. The TN-40 cask is currently..., higher burnup spent fuel used in the PINGP reactor as well as associated changes to the ISFSI's technical...

  5. Detection of West Nile virus-specific antibodies and nucleic acid in horses and mosquitoes, respectively, in Nuevo Leon State, northern Mexico, 2006-2007.

    Science.gov (United States)

    Ibarra-Juarez, L; Eisen, L; Bolling, B G; Beaty, B J; Blitvich, B J; Sanchez-Casas, R M; Ayala-Sulca, Y O; Fernandez-Salas, I

    2012-09-01

    In the last 5 years, there has been only one reported human case of West Nile virus (WNV) disease in northern Mexico. To determine if the virus was still circulating in this region, equine and entomological surveillance for WNV was conducted in the state of Nuevo Leon in northern Mexico in 2006 and 2007. A total of 203 horses were serologically assayed for antibodies to WNV using an epitope-blocking enzyme-linked immunosorbent assay (bELISA). Seroprevalences for WNV in horses sampled in 2006 and 2007 were 26% and 45%, respectively. Mosquito collections in 2007 produced 7365 specimens representing 15 species. Culex mosquitoes were screened for WNV RNA and other genera (Mansonia, Anopheles, Aedes, Psorophora and Uranotaenia) were screened for flaviviruses using reverse-transcription (RT)-PCR. Two pools consisting of Culex spp. mosquitoes contained WNV RNA. Molecular species identification revealed that neither pool included Culex quinquefasciatus (Say) (Diptera:Culicidae) complex mosquitoes. No evidence of flaviviruses was found in the other mosquito genera examined. These data provide evidence that WNV is currently circulating in northern Mexico and that non-Cx. quinquefasciatus spp. mosquitoes may be participating in the WNV transmission cycle in this region. Published 2012. This article is a U.S. Government work and is in the public domain in the USA.

  6. Recent digital control and protection retrofits in power plants

    International Nuclear Information System (INIS)

    Fournier, R.D.; Hammer, M.; Smith, J.E.

    1987-01-01

    Digital computers are now being retrofitted to all types of power plants, replacing analog equipment and solving problems such as equipment obsolescence and low reliability. Three diverse examples of retrofits are presented in this paper, representing trends in man/machine interface design at an oil-fired plant, protection system in pressurized heavy-water reactors, and control systems in light water reactors (LWRs). The examples have been chosen to illustrate diverse reasons for the retrofits and the benefits derived. The cases presented report retrofits at Northern States Power's Monticello boiling water reactor, New Brunswick Electric Power Commission's (NBEPC's) Point Lepreau Nuclear Generating Station, and finally NBEPC's oil-fired plant at Courtney Bay

  7. Northern Dimension: Participant Strategies

    Directory of Open Access Journals (Sweden)

    Busygina Irina

    2009-03-01

    Full Text Available This article is devoted to the “Northern Dimension” initiative of the EU which also includes North-West Russia, Norway and Iceland. It is noted that the “Northern Dimension” in the theoretical perspective can be considered as part of strategic multi-level interactions between member-states of the EU and Russia. On this basis, the authors analyze implications and effects of the strategic interdependence of all the EU-Russia relation levels.

  8. Northern Dimension: Participant Strategies

    Directory of Open Access Journals (Sweden)

    Busygina I.

    2009-01-01

    Full Text Available his article is devoted to the “Northern Dimension” initiative of the EU which also includes North-West Russia, Norway and Iceland. It is noted that the “Northern Dimension” in the theoretical perspective can be considered as part of strategic multi-level interactions between member-states of the EU and Russia. On this basis, the authors analyze implications and effects of the strategic interdependence of all the EU-Russia relation levels.

  9. Factors Affecting Adoption of Agroforestry Farming System as a Mean for Sustainable Agricultural Development and Environment Conservation in Arid Areas of Northern Kordofan State, Sudan

    International Nuclear Information System (INIS)

    Muneer, Siddig El Tayeb

    2008-01-01

    Arid and semi-arid areas represent about 60 percent of Sudan total area. One of the main environmental problems in the arid and semi-arid areas is diffraction's which reduces the natural potential of the already fragile ecosystems and renders rural people vulnerable to food shortages, the vagaries of weather and natural disasters. Deforestation which is considered one of the most critical environmental problems facing the world is one of the main causes of diffraction's. Between the years 1990 and 2005 Sudan lost about 8.8 millions hectares of forests, which represents 11%, of its forests mainly because of subsistence activities such as overgrazing, trees cutting and expansion of traditional agriculture. One of the areas that are very much affected by diffraction's is Northern Kordofan State. To rescue the situation the government of Sudan, with assistance from the United Nations Development Program (UNDP) and some donors, implemented a project that aimed primarily at restocking Acacia Senegal trees in Northern Kordofan State. This study is intended to explore the factors that caused differential rate of farmers' adoption rate of the Acacia Senegal based agroforestry farming system. The study data was collected from a clustered random sample of 300 farmers, through face to face interviews using a questionnaire that was pre-tested and validated. Frequency distribution and multiple regression analysis were used to analyze the data. It has been found that farmers' adoption of agroforestry farming system in Northern Kordofan state was significantly affected by the farmers' level of formal education, contact with extension agents, level of environmental awareness, cosmopoliteness, total area of owned land and extent of social participation. (author)

  10. Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (1)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Sinsuke; Muramatsu, Ken

    2007-02-01

    Japan Atomic Energy Agency (JAEA) entrusted with research on the state-of-the-art of probabilistic safety assessment (PSA) for non-reactor nuclear facilities (NRNF) to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of 'research on the analysis methods for accident consequence in NFRF' was organized in the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. (author)

  11. The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

    International Nuclear Information System (INIS)

    Grundmann, U.

    1995-11-01

    The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

  12. Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (2)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Sinsuke; Muramatsu, Ken

    2007-03-01

    Japan Atomic Energy Agency (JAEA) entrusted with a research on the state-of-the-art of probabilistic safety assessment (PSA) of non-reactor nuclear facilities (NRNF) such as fuel reprocessing and fuel fabrication facilities to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of 'Research on the analysis methods for accident consequence in NFRF' was organized by the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. This report summarizes the results of research conducted by the committee in FY 2005. (author)

  13. An efficient methodology of two groups spatial calculation for neutronic state and sensisivity coefficients in fast reactors

    International Nuclear Information System (INIS)

    Jachic, J.

    1985-01-01

    It is presented the ONEDM neutronic simulator for RZ spatial calculation, two energy groups, aiming at researching and optimization of a low power fast reactor design. The simulator's methodology is based in RZ calculation from radial and axial calculation iteractively coupled and in macroscopic cross sections corrected by power density and asymmetry of the spectrum in the feedback process with phase library for reference neutronic state. The transversal area which are determined by energy groups and material region in the iteration are introduced in the spatial calculation. The simulator efficiency is tested and compared with the CITATION and 2DB codes. The cross sections are generated by 1DX code. (M.C.K.) [pt

  14. Reactor licensing in the United States and Federal Republic of Germany

    International Nuclear Information System (INIS)

    Salvatore, J.E.L.

    1980-02-01

    The licensing procedure for nuclear power plants in the United States and in the Federal Republic of Germany is analysed. The security policy, the inspections and the supervision during their construction and operation are discussed. (A.L.) [pt

  15. Design of the ITER (International Thermonuclear Experimental Reactor) neutral beam system beamline, United States concept

    International Nuclear Information System (INIS)

    Purgalis, P.; Anderson, O.A.; Cooper, W.S.; DeVries, G.E.; Lietzke, A.F.; Kunkel, W.B.; Kwan, J.W.; Matuk, C.A.; Nakai, T.; Stearns, J.W.; Soroka, L.; Wells, R.P.; Lindquist, W.B.; Neef, W.S.; Reginato, L.L.; Sedgley, D.W.; Brook, J.W.; Luzzi, T.E.; Myers, T.J.

    1989-01-01

    Design of a neutral beamline for ITER (International Thermonuclear Experimental Reactor) is described. The design incorporates a barium surface conversion D - source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to watercooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules that can be removed for remote maintenance. The neutral beam system delivers 75 MW of D degree into three ports with a total of nine modules arranged in stacks of three modules per port. To increase reliability each module is designed to deliver up to 10 MW at 1.3 MeV; this allows eight modules operating at partial capacity to deliver the required power in the event one module is removed from service. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 35 m from the port into the torus. Neutron shielding in the drift duct provides the added feature of limiting conductance and thus reducing gas flow to and from the torus. Alternative component choices are also discussed for the evolving design. 8 refs., 4 figs., 1 tab

  16. Status of fast breeder reactor development in the United States of America - April 1984

    International Nuclear Information System (INIS)

    Horton, K.E.

    1984-01-01

    The Breeder Technology program continues to produce viable information on fuel performance, nuclear systems technology, and power conversion technology. The unique testing capabilities design into the FFTF have resulted in well-validated materials and fuels irradiation information that has confirmed and extended previous data bases. Current directions for the research and development program are to improve the technology for power conversion systems, components, instrumentation, and materials technology to the point where cost reduction and reliability potentials are realized. Operation of the breeder test facility complex at the Hanford Engineering Development Laboratory (HEDL), the Energy Technology Engineering Center (ETEC), and the Argonne National Laboratory (ANL) continues to provide the experience base and test capability for the breeder R and D effort. International cooperation will be even more important in the future than in the past for several reasons. Significant new investments still have to be made in breeder R and D to improve designs, achieve economic competitiveness and to develop practical breeder fuel cycle capabilities. Progress can be accelerated, redundancies avoided, and economics achieved if nations coordinate their programs, and where possible, divide up the work. In addition, there is clear mutual benefit in encouraging the countries involved in breeder development to harmonize standards and regulations related to safety. It is also important that the advanced nations work together closely in assuring that adequate international safeguards, export controls, and national physical security measures keep pace with breeder reactor and fuel cycle developments

  17. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  18. Rotating reactors : a review

    NARCIS (Netherlands)

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  19. An innovation for improving maternal, newborn and child health (MNCH) service delivery in Jigawa State, northern Nigeria: a qualitative study of stakeholders' perceptions about clinical mentoring.

    Science.gov (United States)

    Okereke, Ekechi; Tukur, Jamilu; Aminu, Amina; Butera, Jean; Mohammed, Bello; Tanko, Mustapha; Yisa, Ibrahim; Obonyo, Benson; Egboh, Mike

    2015-02-15

    An effective capacity building process for healthcare workers is required for the delivery of quality health care services. Work-based training can be applied for the capacity building of health care workers while causing minimum disruption to service delivery within health facilities. In 2012, clinical mentoring was introduced into the Jigawa State Health System through collaboration between the Jigawa State Ministry of Health and the Partnership for Transforming Health Systems Phase 2 (PATHS2). This study evaluates the perceptions of different stakeholders about clinical mentoring as a strategy for improving maternal, newborn and child health service delivery in Jigawa State, northern Nigeria. Interviews were conducted in February 2013 with different stakeholders within Jigawa State in Northern Nigeria. There were semi-structured interviews with 33 mentored health care workers as well as the health facility departmental heads for Obstetrics and Pediatrics in the selected clinical mentoring health facilities. In-depth interviews were also conducted with the clinical mentors and two senior government health officials working within the Jigawa State Ministry of Health. The qualitative data were audio-recorded; transcribed and thematically analysed. The study findings suggest that clinical mentoring improved service delivery within the clinical mentoring health facilities. Significant improvements in the professional capacity of mentored health workers were observed by clinical mentors, heads of departments and the mentored health workers. Best practices were introduced with the support of the clinical mentors such as appropriate baseline investigations for pediatric patients, the use of magnesium sulphate and misoprostol for the management of eclampsia and post-partum hemorrhage respectively. Government health officials indicate that clinical mentoring has led to more emphasis on the need for the provision of better quality health services. Stakeholders report that

  20. State of radioactive waste management is power reactor facilities and state of radiation exposure of workers who engaged in radiation works in fiscal 1993

    International Nuclear Information System (INIS)

    1994-01-01

    This report is the summary of the reports on radiation control and others submitted by those who installed practical power reactor facilities based on the relevant law in fiscal 1993. The amounts of release of radioactive gaseous and liquid wastes were sufficiently smaller than the target value of the yearly release control for attaining the target value of dose that the public around the facilities receive. As to the state of control of radioactive solid waste, the amount of drum generation tended to decrease year by year, and the cumulative amount to be preserved tended to level off. The dose equivalent that the individuals who engaged in radiation works received was smaller than the limit value in all nuclear power stations. The total dose equivalent for those workers in fiscal 1993 was 86.65 man Sv. Hereafter, the automation and remote operation of works, the water quality control for reducing crud and so on will be promoted to reduce radiation exposure. The reference data on the state of control of gaseous, liquid and solid wastes, and the state of control of radiation exposure of workers are attached. (K.I.)

  1. Prevalence of HCV infection and associated factors among illicit drug users in Breves, State of Pará, northern Brazil

    OpenAIRE

    Pacheco,Suzy Danielly Barbosa; Silva-Oliveira,Gláucia Caroline; Maradei-Pereira,Luciana Maria Cunha; Crescente,José Ângelo Barletta; Lemos,José Alexandre Rodrigues de; Oliveira-Filho,Aldemir Branco de

    2014-01-01

    Introduction: Illicit drug users (DUs) are vulnerable to hepatitis C virus (HCV) infection. The shared use of illicit drugs is the main method of HCV transmission. Methods: A cross-sectional study was conducted in Breves, in northern Brazil. We surveyed 187 DUs to determine the prevalence of and factors associated with HCV infection. Results: The prevalence of anti-HCV antibodies was 36.9%, and the prevalence of hepatitis C virus-ribonucleic acid (HCV-RNA) was 31%. Hepatitis C virus infec...

  2. Radioactivity computation of steady-state and pulsed fusion reactors operation

    International Nuclear Information System (INIS)

    Attaya, H.

    1994-06-01

    Different mathematical methods are used to calculate the nuclear transmutation in steady-state and pulsed neutron irradiation. These methods are the Schuer decomposition, the eigenvector decomposition, and the Pade approximation of the matrix exponential function. In the case of the linear decay chain approximation, a simple algorithm is used to evaluate the transition matrices

  3. Nuclear reactors built, being built, or planned in the United States

    International Nuclear Information System (INIS)

    1984-08-01

    This booklet contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of Apr. 30, 1984, which are capable of sustaining a nuclear chain reaction. Information is presented in five parts, each of which is categorized by primary function or purpose: civilian, military, production, export, and critical assembly facilities

  4. Nuclear reactors built, being built, or planned in the United States

    International Nuclear Information System (INIS)

    Carter, E.P.

    1985-09-01

    This publication contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of March 1985, which are capable of sustaining a nuclear chain reaction. Information is presented in five parts, each of which is categorized by primary function or purpose: civilian, production, military, export, and critical assembly facilities

  5. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  6. Steady-state CFD simulations of an EPR™ reactor pressure vessel: A validation study based on the JULIETTE experiments

    International Nuclear Information System (INIS)

    Puragliesi, R.; Zhou, L.; Zerkak, O.; Pautz, A.

    2016-01-01

    Highlights: • CFD validation of k–ε (RANS model of EPR RPV. • Flat inlet velocity profile is not sufficient to correctly predict the pressure drops. • Swirl is responsible for asymmetric loads at the core barrel. • Parametric study to the turbulent Schmidt number for better predictions of passive-scalar transport. • The optimal turbulent Schmidt number was found to be one order of magnitude smaller than the standard value. - Abstract: Validating computational fluid dynamics (CFD) models against experimental measurements is a fundamental step towards a broader acceptance of CFD as a tool for reactor safety analysis when best-estimate one-dimensional thermal-hydraulic codes present strong modelling limitations. In the present paper numerical results of steady-state RANS analyses are compared to pressure, volumetric flow rate and concentration distribution measurements in different locations of an Areva EPR™ reactor pressure vessel (RPV) mock-up named JULIETTE. Several flow configurations are considered: Three different total volumetric flow rates, cold leg velocity field with or without swirl, three or four reactor coolant pumps functioning. Investigations on the influence of two types of inlet boundary profiles (i.e. flat or 1/7th power-law) and the turbulent Schmidt number have shown that the first affects sensibly the pressure loads at the core barrel whereas the latter parameter strongly affects the transport and the mixing of the tracer (passive scalar) and consequently its distribution at the core inlet. Furthermore, the introduction of an integral parameter as the swirl number has helped to decrease the large epistemic uncertainty associated with the swirling device. The swirl is found to be the cause of asymmetric loads on the walls of the core barrel and also asymmetries are enhanced for the tracer concentration distribution at the core inlet. The k–ϵ CFD model developed with the commercial code STAR-CCM+ proves to be able to predict

  7. Steady-state CFD simulations of an EPR™ reactor pressure vessel: A validation study based on the JULIETTE experiments

    Energy Technology Data Exchange (ETDEWEB)

    Puragliesi, R., E-mail: riccardo.puragliesi@psi.ch [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland); Zhou, L. [Science and Technology on Reactor System Design Technology Laboratory, NPIC, Chengdu (China); Zerkak, O.; Pautz, A. [Laboratory for Reactor Physics and Systems Behaviour, PSI, 5232 Villigen (Switzerland)

    2016-04-15

    Highlights: • CFD validation of k–ε (RANS model of EPR RPV. • Flat inlet velocity profile is not sufficient to correctly predict the pressure drops. • Swirl is responsible for asymmetric loads at the core barrel. • Parametric study to the turbulent Schmidt number for better predictions of passive-scalar transport. • The optimal turbulent Schmidt number was found to be one order of magnitude smaller than the standard value. - Abstract: Validating computational fluid dynamics (CFD) models against experimental measurements is a fundamental step towards a broader acceptance of CFD as a tool for reactor safety analysis when best-estimate one-dimensional thermal-hydraulic codes present strong modelling limitations. In the present paper numerical results of steady-state RANS analyses are compared to pressure, volumetric flow rate and concentration distribution measurements in different locations of an Areva EPR™ reactor pressure vessel (RPV) mock-up named JULIETTE. Several flow configurations are considered: Three different total volumetric flow rates, cold leg velocity field with or without swirl, three or four reactor coolant pumps functioning. Investigations on the influence of two types of inlet boundary profiles (i.e. flat or 1/7th power-law) and the turbulent Schmidt number have shown that the first affects sensibly the pressure loads at the core barrel whereas the latter parameter strongly affects the transport and the mixing of the tracer (passive scalar) and consequently its distribution at the core inlet. Furthermore, the introduction of an integral parameter as the swirl number has helped to decrease the large epistemic uncertainty associated with the swirling device. The swirl is found to be the cause of asymmetric loads on the walls of the core barrel and also asymmetries are enhanced for the tracer concentration distribution at the core inlet. The k–ϵ CFD model developed with the commercial code STAR-CCM+ proves to be able to predict

  8. Analysis of reactor power behaviour using estimation of period for the gain adaptation in a state feedback controller; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Benitez R, J.S. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Perez C, J.H. [CINVESTAV, IPN, A.P. 14740 07000 Mexico D.F. (Mexico); Rivero G, T. [ITT, 50140 Metepec, Estado de Mexico (Mexico)

    2008-07-01

    In this paper a novel procedure for power regulation in a TRIGA Mark III nuclear reactor is presented. The control scheme combines state variable feedback with a first order predictor, which is incorporated to speed up the power response of the reactor without exceeding the safety requirement imposed by the reactor period. The simulation results using the proposed control strategy attains different values of steady-state power from different values of initial power in short time, complying at all times with the safety restriction imposed on the reactor period. The predictor, derived from the theory of first order numerical integration, produces very good results during the ascent of power. These results include a fast response and independence of the wide variety of potential operating conditions something not easy and even impossible to obtain with other procedures. By using this control scheme, the reactor period is maintained within safety limits during the start up of the reactor, which is normally the operating condition where an occurrence of a period scram is common. However, the predictor can not be used when the power is reaching the desired power level because the instantaneous power increases far above the desired level. Thus, when the power increases above certain power level, the state feedback gain is set constant to a predefined value. This causes some oscillations that decrease in a few seconds. Afterwards, the power response smoothly approaches, with a small overshoot, the desired power. This constraint on the use of the predictor prevents the unbounded increase of the neutron power. The control law proposed requires all the system's state variables. Since only the neutron power is available, it is necessary the estimation of the non measurable states. The key issue of the existence of a solution to this problem has been previously considered. One of the conclusions is that the point kinetic equations are observable under certain restrictions

  9. Development of a practical Monte Carlo based fuel management system for the Penn State University Breazeale Research Reactor (PSBR)

    International Nuclear Information System (INIS)

    Tippayakul, Chanatip; Ivanov, Kostadin; Frederick Sears, C.

    2008-01-01

    A practical fuel management system for the he Pennsylvania State University Breazeale Research Reactor (PSBR) based on the advanced Monte Carlo methodology was developed from the existing fuel management tool in this research. Several modeling improvements were implemented to the old system. The improved fuel management system can now utilize the burnup dependent cross section libraries generated specifically for PSBR fuel and it is also able to update the cross sections of these libraries by the Monte Carlo calculation automatically. Considerations were given to balance the computation time and the accuracy of the cross section update. Thus, certain types of a limited number of isotopes, which are considered 'important', are calculated and updated by the scheme. Moreover, the depletion algorithm of the existing fuel management tool was replaced from the predictor only to the predictor-corrector depletion scheme to account for burnup spectrum changes during the burnup step more accurately. An intermediate verification of the fuel management system was performed to assess the correctness of the newly implemented schemes against HELIOS. It was found that the agreement of both codes is good when the same energy released per fission (Q values) is used. Furthermore, to be able to model the reactor at various temperatures, the fuel management tool is able to utilize automatically the continuous cross sections generated at different temperatures. Other additional useful capabilities were also added to the fuel management tool to make it easy to use and be practical. As part of the development, a hybrid nodal diffusion/Monte Carlo calculation was devised to speed up the Monte Carlo calculation by providing more converged initial source distribution for the Monte Carlo calculation from the nodal diffusion calculation. Finally, the fuel management system was validated against the measured data using several actual PSBR core loadings. The agreement of the predicted core

  10. Northern employment

    International Nuclear Information System (INIS)

    Zavitz, J.

    1997-01-01

    Hiring practices and policies and employment opportunities that were available in the Beaufort Sea and MacKenzie Delta project for local residents and for people from southern Canada were dealt with in this chapter. Depending on the source, Northern hiring was a mere token, or a genuine and successful effort on the part of the companies to involve the native population and to share with them the benefits of the project. The fact remains that opening up job opportunities for Northerners was not easily attained, and would never have been realized without the involvement of government and community organizations. Government also played a major role in developing policies and training regimes. By the end of exploration operations, the hiring of Northern residents in the oil and gas industry had become a requirement of drilling applications. Training programs were also created to ensure that Northern residents received the means necessary to take advantage of Northern employment opportunities

  11. Financing the next generation of new reactors in the united states. Panel Discussion

    International Nuclear Information System (INIS)

    Turner, Kyle; Simard, Ron; Tran, K.C.; Kelly, Patrick; Green, Barrett E.; Quinn, Edward L.; Stamos, John

    2001-01-01

    Full text of publication follows: With the California energy shortage and new growth forecasts in the United States, significant new base-load generation will be needed in the near future to meet electricity demands. New figures for growth in electricity demand for the United States rose significantly because of Internet and related business expansion. Lack of sufficient natural gas supplies to support new generation in some regions is causing a renewed interest in building new nuclear plants. Speakers will address the current status of available and near-term design options including both the U.S. Department of Energy Generation III and IV design packages, infrastructure challenges, and financial models that show that nuclear is competitive with alternatives and a prudent and profitable investment. (authors)

  12. Evaluation of Productivity of Zymotis Solid-State Bioreactor Based on Total Reactor Volume

    Directory of Open Access Journals (Sweden)

    Oscar F. von Meien

    2002-01-01

    Full Text Available In this work a method of analyzing the performance of solid-state fermentation bioreactors is described. The method is used to investigate the optimal value for the spacing between the cooling plates of the Zymotis bioreactor, using simulated fermentation data supplied by a mathematical model. The Zymotis bioreactor has good potential for those solid-state fermentation processes in which the substrate bed must remain static. The current work addresses two design parameters introduced by the presence of the internal heat transfer plates: the width of the heat transfer plate, which is governed by the amount of heat to be removed and the pressure drop of the cooling water, and the spacing between these heat transfer plates. In order to analyze the performance of the bioreactor a productivity term is introduced that takes into account the volume occupied within the bioreactor by the heat transfer plates. As part of this analysis, it is shown that, for logistic growth kinetics, the time at which the biomass reaches 90 % of its maximum possible value is a good estimate of the optimum harvesting time for maximizing productivity. Application of the productivity analysis to the simulated fermentation results suggests that, with typical fast growing fungi ( = 0.324 h–1, the optimal spacing between heat transfer plates is of the order of 6 cm. The general applicability of this approach to evaluate the productivity of solid-state bioreactors is demonstrated.

  13. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  14. Regional Characteristics of Stress State of Main Seismic Active Faults in Mid-Northern Part of Sichuan-Yunnan Block

    Science.gov (United States)

    Weiwei, W.; Yaling, W.

    2017-12-01

    We restore the seismic source spectrums of 1012 earthquakes(2.0 ≤ ML ≤ 5.0) in the mid-northern part of Sichuan-Yunnan seismic block(26 ° N-33 ° N, 99 ° E-104 ° E),then calculate the source parameters.Based on the regional seismic tectonic background, the distribution of active faults and seismicity, the study area is divided into four statistical units (Z1 Jinshajiang and Litang fault zone, Z2 Xianshuihe fault zone, Z3 Anninghe-Zemuhe fault zone, Z4 Lijiang-Xiaojinhe fault zone). Seismic source stress drop results show the following, (1)The stress at the end of the Jinshajiang fault is low, strong earthquake activity rare.Stress-strain loading deceases gradually from northwest to southeast along Litang fault, the northwest section which is relatively locked is more likely to accumulate strain than southeast section. (2)Stress drop of Z2 is divided by Kangding, the southern section is low and northern section is high. Southern section (Kangding-Shimian) is difficult to accumulate higher strain in the short term, but in northern section (Garzê-Kangding), moderate and strong earthquakes have not filled the gaps of seismic moment release, there is still a high stress accumulation in partial section. (3)High stress-drop events were concentrated on Z3, strain accumulation of this unit is strong, and stress level is the highest, earthquake risk is high. (4)On Z4, stress drop characteristics of different magnitude earthquakes are not the same, which is related to complex tectonic setting, the specific reasons still need to be discussed deeply.The study also show that, (1)Stress drops display a systematic change with different faults and locations, high stress-drop events occurs mostly on the fault intersection area. Faults without locking condition and mainly creep, are mainly characterized by low stress drop. (2)Contrasting to what is commonly thought that "strike-slip faults are not easy to accumulate stress ", Z2 and Z3 all exhibit high stress levels, which

  15. State of the art of computer codes and experimental investigations on safety of nuclear power plants with reactors of WWER type

    International Nuclear Information System (INIS)

    Asmolov, V.G.; Volkov, G.A.; Elikin, I.V.; Mysenkov, A.I.

    1987-01-01

    As is well-known investigations on mathematical models of accidental situations from the point of view of nuclear safety as well as their experimental investigation are of great importance in the design of reactor units for nuclear power plants. This paper gives a review of the state of the art of thermodynamic models and computer codes used for safety analysis of WWER reactors in the USSR, the experimental basis and experimental investigations of the appropriate thermal processes. The actual and future trends of theoretical and experimental investigation on safety problems of WWER type nuclear power plants are briefly described. (author)

  16. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  17. The behaviour of water-cooled reactor fuel rods in steady state and transient conditions

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.

    1997-01-01

    In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author)

  18. Epithermal neutron beam for BNCT research at the Washington State University TRIGA research reactor

    International Nuclear Information System (INIS)

    Nigg, D.W.; Venhuizen, J.R.; Wheeler, F.J.; Wemple, C.A.; Tripard, G.E.; Gavin, P.R.

    2000-01-01

    A new epithermal-neutron beam facility for BNCT (Boron Neutron Capture Therapy) research and boronated agent screening in animal models is in the final stages of construction at Washington State University (WSU). A key distinguishing feature of the design is the incorporation of a new, high-efficiency, neutron moderating and filtering material, Fluental, developed by the Technical Research Centre of Finland. An additional key feature is the provision for adjustable filter-moderator thickness to systematically explore the radiobiological consequences of increasing the fast-neutron contamination above the nominal value associated with the baseline system. (author)

  19. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  20. Core-state models for fuel management of equilibrium and transition cycles in pressurized water reactors

    International Nuclear Information System (INIS)

    Aragones, J.M.; Martinez-Val, J.M.; Corella, M.R.

    1977-01-01

    Fuel management requires that mass, energy, and reactivity balance be satisfied in each reload cycle. Procedures for selection of alternatives, core-state models, and fuel cost calculations have been developed for both equilibrium and transition cycles. Effective cycle lengths and fuel cycle variables--namely, reload batch size, schedule of incore residence for the fuel, feed enrichments, energy sharing cycle by cycle, and discharge burnup and isotopics--are the variables being considered for fuel management planning with a given energy generation plan, fuel design, recycling strategy, and financial assumptions