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Sample records for fuel project dose

  1. Spent Nuclear Fuel Project dose management plan

    International Nuclear Information System (INIS)

    Bergsman, K.H.

    1996-03-01

    This dose management plan facilitates meeting the dose management and ALARA requirements applicable to the design activities of the Spent Nuclear Fuel Project, and establishes consistency of information used by multiple subprojects in ALARA evaluations. The method for meeting the ALARA requirements applicable to facility designs involves two components. The first is each Spent Nuclear Fuel Project subproject incorporating ALARA principles, ALARA design optimizations, and ALARA design reviews throughout the design of facilities and equipment. The second component is the Spent Nuclear Fuel Project management providing overall dose management guidance to the subprojects and oversight of the subproject dose management efforts

  2. Fuel cells: Project Volta

    Energy Technology Data Exchange (ETDEWEB)

    Vellone, R.; Di Mario, F.

    1987-09-01

    This paper discusses research and development in the field of fuel cell power plants. Reference is made to the Italian research Project Volta. Problems related to research program financing and fuel cell power plant marketing are discussed.

  3. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  4. Severe fuel damage projects

    International Nuclear Information System (INIS)

    Sdouz, G.

    1987-10-01

    After the descriptions of the generation of a Severe Fuel Damage Accident in a LWR the hypothetical course of such an accident is explained. Then the most significant projects are described. At each project the experimental facility, the most important results and the concluding models and codes are discussed. The selection of the projects is concentrated on the German Projekt Nukleare Sicherheit (PNS), tests performed at the Idaho National Engineering Laboratory (INEL) and smaller projects in France and Great Britain. 25 refs., 26 figs. (Author)

  5. Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Cannon, S.D.; Finch, S.M.

    1992-10-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation doses that individuals and populations could have received from nuclear operations at Hanford since 1944. The independent Technical Steering Panel (TSP) provides technical direction. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed from release to impact on humans (dose estimates):Source Terms, Environmental Transport, Environmental Monitoring Data, Demography, Food Consumption, and Agriculture, and Environmental Pathways and Dose Estimates

  6. Radkowsky Thorium Fuel Project

    International Nuclear Information System (INIS)

    Todosow, Michael

    2006-01-01

    In the early/mid 1990's Prof. Alvin Radkowsky, former chief scientist of the U.S. Naval Reactors program, proposed an alternate fuel concept employing thorium-based fuel for use in existing/next generation pressurized water reactors (PWRs). The concept was based on the use of a 'seed-blanket-unit' (SBU) that was a one-for-one replacement for a standard PWR assembly with a uranium-based central 'driver' zone, surrounded by a 'blanket' zone containing uranium and thorium. Therefore, the SBU could be retrofit without significant modifications into existing/next generation PWRs. The objective was to improve the proliferation and waste characteristics of the current once-through fuel cycle. The objective of a series of projects funded by the Initiatives for Proliferation Prevention program of the U.S. Department of Energy (DOE-IPP) - BNL-T2-0074,a,b-RU 'Radkowsky Thorium Fuel (RTF) Concept' - was to explore the characteristics and potential of this concept. The work was performed under several BNL CRADAs (BNL-C-96-02 and BNL-C-98-15) with the Radkowsky Thorium Power Corp./Thorium Power Inc. and utilized the technical and experimental capabilities in the Former Soviet Union (FSU) to explore the potential of this concept for implementation in Russian pressurized water reactors (VVERs), and where possible, also generate data that could be used for design and licensing of the concept for Western PWRs. The Project in Russia was managed by the Russian Research Center-?'Kurchatov Institute' (RRC-KI), and included several institutes (e.g., PJSC 'Electrostal', NPO 'LUCH' (Podolsk), RIINM (Bochvar Institute), GAN RF (Gosatomnadzor), Kalininskaja NPP (VVER-1000)), and consisted of the following phases: Phase-1 ($550K/$275K to Russia): The objective was to perform an initial review of all aspects of the concept (design, performance, safety, implementation issues, cost, etc.) to confirm feasibility/viability and identify any 'show-stoppers'; Phase-2 ($600K/$300K to Russia

  7. Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Finch, S.M.; McMakin, A.H.

    1991-04-01

    The objective of the Hanford Environmental Dose Reconstruction Project is to estimate the radiation doses that populations could have received from nuclear operations at Hanford since 1944. The project is being managed and conducted by the Pacific Northwest Laboratory (PNL) under the direction of an independent Technical Steering Panel (TSP). The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed, from released to impact on humans (dose estimates): source terms; environmental transport; environmental monitoring data; demographics, agriculture, food habits; and, environmental pathways and dose estimates

  8. Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Finch, S.M.; McMakin, A.H.

    1992-06-01

    The objective of the Hanford Environmental Dose Reconstruction Project is to estimate the radiation doses that individuals and populations could have received from nuclear operations at Hanford since 1944. The project is being managed and conducted by the Battelle Pacific Northwest Laboratories under contract with the Centers for Disease Control. The independent Technical Steering Panel (TSP) provides technical direction. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed, from release to impact on humans (dose estimates): source terms; environmental transport; environmental monitoring data; demography, food consumption, and agriculture; environmental pathways and dose estimates

  9. Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    McMakin, A.H.; Cannon, S.D.; Finch, S.M.

    1992-07-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation doses that individuals and populations could have received from nuclear operations at Hanford since 1944. The TSP consists of experts in environmental pathways, epidemiology, surface-water transport, ground-water transport, statistics, demography, agriculture, meteorology, nuclear engineering, radiation dosimetry, and cultural anthropology. Included are appointed technical members representing the states of Oregon, Washington, and Idaho, a representative of Native American tribes, and an individual representing the public. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed from release to impact on humans (dose estimates): Source terms, environmental transport, environmental monitoring data, demography, food consumption, and agriculture, and environmental pathways and dose estimates. Progress is discussed

  10. Project fuel development

    International Nuclear Information System (INIS)

    Stratton, R.W.

    1981-05-01

    The activities continued on lab-scale production of uranium-plutonium carbide fuel for the fast reactor using gelation methods, irradiation testing and performance evaluation. Whereas in earlier years a balance was maintained between research and development or with emphasis on research, 1980 was marked by a concentrated equipment development effort for an increased throughput with improved process control and product reproducability and installation of new equipment for large pin fabrication. (Auth.)

  11. Spent Nuclear Fuel Project Safety Management Plan

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities

  12. Canadian fusion fuels technology project

    International Nuclear Information System (INIS)

    1986-01-01

    The Canadian Fusion Fuels Technology Project was launched in 1982 to coordinate Canada's provision of fusion fuels technology to international fusion power development programs. The project has a mandate to extend and adapt existing Canadian tritium technologies for use in international fusion power development programs. 1985-86 represents the fourth year of the first five-year term of the Canadian Fusion Fuels Technology Project (CFFTP). This reporting period coincides with an increasing trend in global fusion R and D to direct more effort towards the management of tritium. This has resulted in an increased linking of CFFTP activities and objectives with those of facilities abroad. In this way there has been a continuing achievement resulting from CFFTP efforts to have cooperative R and D and service activities with organizations abroad. All of this is aided by the cooperative international atmosphere within the fusion community. This report summarizes our past year and provides some highlights of the upcoming year 1986/87, which is the final year of the first five-year phase of the program. AECL (representing the Federal Government), the Ministry of Energy (representing Ontario) and Ontario Hydro, have given formal indication of their intent to continue with a second five-year program. Plans for the second phase will continue to emphasize tritium technology and remote handling

  13. Hanford Environmental Dose Reconstruction Project monthly report

    International Nuclear Information System (INIS)

    Finch, S.M.

    1990-12-01

    The objective of the Hanford Environmental Dose Reconstruction Project is to estimate the radiation doses that populations could have been have received from nuclear operations at Hanford since 1944. The project is being managed and conducted by the Pacific Northwest Laboratory (PNL) under the direction of an independent Technical Steering Panel (TSP). The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed, from release to impact on humans (dose estimates): source terms; environmental transport; environmental monitoring data; demographics, agriculture, food habits; and environmental pathways and dose estimates. 3 figs., 3 tabs

  14. Hanford Environmental Dose Reconstruction Project monthly report

    International Nuclear Information System (INIS)

    Finch, S.M.

    1991-10-01

    The objective of the Hanford Environmental Dose Reconstruction Project is to estimate the radiation doeses that individuals and populations could have received from nuclear operations at Hanford since 1944. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed, from release to impact on humans (dose estimates): Source terms; environmental transport; environmental monitoring data; demographics, agriculture, food habits; environmental pathways and dose estimates

  15. Navy fuel cell demonstration project.

    Energy Technology Data Exchange (ETDEWEB)

    Black, Billy D.; Akhil, Abbas Ali

    2008-08-01

    This is the final report on a field evaluation by the Department of the Navy of twenty 5-kW PEM fuel cells carried out during 2004 and 2005 at five Navy sites located in New York, California, and Hawaii. The key objective of the effort was to obtain an engineering assessment of their military applications. Particular issues of interest were fuel cell cost, performance, reliability, and the readiness of commercial fuel cells for use as a standalone (grid-independent) power option. Two corollary objectives of the demonstration were to promote technological advances and to improve fuel performance and reliability. From a cost perspective, the capital cost of PEM fuel cells at this stage of their development is high compared to other power generation technologies. Sandia National Laboratories technical recommendation to the Navy is to remain involved in evaluating successive generations of this technology, particularly in locations with greater environmental extremes, and it encourages their increased use by the Navy.

  16. The Hanford Environmental Dose Reconstruction Project: Overview

    International Nuclear Information System (INIS)

    Haerer, H.A.; Freshley, M.D.; Gilbert, R.O.; Morgan, L.G.; Napier, B.A.; Rhoads, R.E.; Woodruff, R.K.

    1990-01-01

    In 1988, researchers began a multiyear effort to estimate radiation doses that people could have received since 1944 at the U.S. Department of Energy's Hanford Site. The study was prompted by increasing concern about potential health effects to the public from more than 40 yr of nuclear activities. We will provide an overview of the Hanford Environmental Dose Reconstruction Project and its technical approach. The work has required development of new methods and tools for dealing with unique technical and communication challenges. Scientists are using a probabilistic, rather than the more typical deterministic, approach to generate dose distributions rather than single-point estimates. Uncertainties in input parameters are reflected in dose results. Sensitivity analyses are used to optimize project resources and define the project's scope. An independent technical steering panel directs and approves the work in a public forum. Dose estimates are based on review and analysis of historical data related to operations, effluents, and monitoring; determination of important radionuclides; and reconstruction of source terms, environmental conditions that affected transport, concentrations in environmental media, and human elements, such as population distribution, agricultural practices, food consumption patterns, and lifestyles. A companion paper in this volume, The Hanford Environmental Dose Reconstruction Project: Technical Approach, describes the computational framework for the work

  17. PBMR Project - Pebble Fuel Advantages

    International Nuclear Information System (INIS)

    Slabber, Johan; Matzie, Regis; Casperson, Sten; Kriel, Willem

    2006-01-01

    An overview is presented of all the important issues that influenced the choice of pebble fuel for the High-temperature Gas-cooled Reactor (HTGR) concept developed by South Africa. Each of these issues is then discussed in detail and compared with other fuel configurations proposed for direct cycle High-temperature Reactor (HTR) applications. The comparisons are provided using objective data generated by analyses done for the design of the Pebble Bed Modular Reactor (PBMR) and data that is available in open literature for the other fuel configurations

  18. Hanford Environmental Dose Reconstruction Project monthly report

    International Nuclear Information System (INIS)

    McMakin, A.H., Cannon, S.D.; Finch, S.M.

    1992-09-01

    The objective of the Hanford Environmental Dose Reconstruction MDR) Project is to estimate the radiation doses that individuals and populations could have received from nuclear operations at Hanford since 1944. The TSP consists of experts in envirorunental pathways. epidemiology, surface-water transport, ground-water transport, statistics, demography, agriculture, meteorology, nuclear engineering. radiation dosimetry. and cultural anthropology. Included are appointed members representing the states of Oregon, Washington, and Idaho, a representative of Native American tribes, and an individual representing the public. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed from release to impact on humans (dose estimates): Source Terms; Environmental Transport; Environmental Monitoring Data Demography, Food Consumption, and Agriculture; and Environmental Pathways and Dose Estimates

  19. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    LEROY, P.G.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  20. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  1. Maritime Fuel Cell Generator Project.

    Energy Technology Data Exchange (ETDEWEB)

    Pratt, Joseph William [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2017-07-01

    Fuel costs and emissions in maritime ports are an opportunity for transportation energy efficiency improvement and emissions reduction efforts. Ocean-going vessels, harbor craft, and cargo handling equipment are still major contributors to air pollution in and around ports. Diesel engine costs continually increase as tighter criteria pollutant regulations come into effect and will continue to do so with expected introduction of carbon emission regulations. Diesel fuel costs will also continue to rise as requirements for cleaner fuels are imposed. Both aspects will increase the cost of diesel-based power generation on the vessel and on shore. Although fuel cells have been used in many successful applications, they have not been technically or commercially validated in the port environment. One opportunity to do so was identified in Honolulu Harbor at the Young Brothers Ltd. wharf. At this facility, barges sail regularly to and from neighbor islands and containerized diesel generators provide power for the reefers while on the dock and on the barge during transport, nearly always at part load. Due to inherent efficiency characteristics of fuel cells and diesel generators, switching to a hydrogen fuel cell power generator was found to have potential emissions and cost savings.

  2. Spent Nuclear Fuel Project operational staffing plan

    International Nuclear Information System (INIS)

    Debban, B.L.

    1996-03-01

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  3. Thermodynamics of Advanced Fuels - International Database Project

    International Nuclear Information System (INIS)

    Massara, Simone; Gueneau, Christine

    2014-01-01

    The Thermodynamics of Advanced Fuels - International Database (TAF-ID) Project was established in 2013 under the auspices of the NEA Nuclear Science Committee. The project was designed to make available a comprehensive, internationally recognised and quality-assured database of phase diagrams and thermodynamic properties of advanced nuclear fuels with a view to meeting specialised requirements for the development of advanced fuels for a future generation of nuclear reactors. Some of the specific technical objectives that this programme intends to achieve are to predict the solid, liquid and/or gas phases formed during fuel cladding chemical interactions under normal and accident conditions, to improve the control of the experimental conditions during the fabrication of fuel materials at high temperature, for example by predicting the vapour pressures of the elements (particularly of plutonium and the minor actinides) and to predict the evolution of the chemical composition of fuel under irradiation versus temperature and burn-up. This joint project, co-ordinated by the NEA, was established for an initial three-year period among nine organisations from six NEA member countries: Canada (AECL, RMCC, UOIT), France (CEA), Japan (JAEA, CRIEPI), the Netherlands (NRG), the Republic of Korea (KAERI) and the United States (US DOE). It is entirely funded by the nine signatories of the project. (authors)

  4. The EFR project: core and fuel

    International Nuclear Information System (INIS)

    Francillon, E.; Barnes, D.W.; Pay, A.; Wehmann, U.

    1991-01-01

    The draft studies on EFR has beginning, in March 1988. The first part of the summary draft has consisted in the core and fuel domains to harmonize the different approaches used in national projects (SPX2-SNR2-CDFR). Rapidly, the core First Consistent Design has been defined with references to the anterior conceptions. Since this definition, studies have been engaged on the management (mean burnup amelioration) and on the conception (breeding gain, sodium void coefficient reduction). After a presentation of the basis options and on the general conception of the fuel assemblies we make a point on the core and the interfaces with the fuel cycle [fr

  5. Employee dose reduction at British Nuclear Fuels plc

    International Nuclear Information System (INIS)

    Fishwick, A.H.; Finlayson, J.L.; James, R.D.

    1992-01-01

    Average work force doses in uranium fuel fabrication plants are a small percentage (about 6 % or 3 mSv pa) of UK regulatory limits. In uranium metal casting, and uranium oxide production plants, doses are somewhat higher than the average. Dose reduction methods have, however, resulted in these being reduced to 20 %, or less, of the same limit. Major future investment should reduce doses in oxide production plants to about the current average level. (author)

  6. Diesel fueled ship propulsion fuel cell demonstration project

    Energy Technology Data Exchange (ETDEWEB)

    Kumm, W.H. [Arctic Energies Ltd., Severna Park, MD (United States)

    1996-12-31

    The paper describes the work underway to adapt a former US Navy diesel electric drive ship as a 2.4 Megawatt fuel cell powered, US Coast Guard operated, demonstrator. The Project will design the new configuration, and then remove the four 600 kW diesel electric generators and auxiliaries. It will design, build and install fourteen or more nominal 180 kW diesel fueled molten carbonate internal reforming direct fuel cells (DFCs). The USCG cutter VINDICATOR has been chosen. The adaptation will be carried out at the USCG shipyard at Curtis Bay, MD. A multi-agency (state and federal) cooperative project is now underway. The USCG prime contractor, AEL, is performing the work under a Phase III Small Business Innovation Research (SBIR) award. This follows their successful completion of Phases I and II under contract to the US Naval Sea Systems (NAVSEA) from 1989 through 1993 which successfully demonstrated the feasibility of diesel fueled DFCs. The demonstrated marine propulsion of a USCG cutter will lead to commercial, naval ship and submarine applications as well as on-land applications such as diesel fueled locomotives.

  7. Spent fuel and waste inventories and projections

    International Nuclear Information System (INIS)

    Carter, W.L.; Finney, B.C.; Alexander, C.W.; Blomeke, J.O.; McNair, J.M.

    1980-08-01

    Current inventories of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on judgments of the most reliable information available from Government sources and the open literature. Future waste generation rates and quantities to be accumulated over the remainder of this century are also presented, based on a present projection of US commercial nuclear power growth and expected defense-related activities. Spent fuel projections are based on the current DOE/EIA estimate of nuclear growth, which projects 180 GW(e) in the year 2000. It is recognized that the calculated spent fuel discharges are probably high in view of recent reactor cancellations; hence adjustments will be made in future updates of this report. Wastes considered, on a chapter-by-chapter basis, are: spent fuel, high-level wastes, transuranic wastes, low-level wastes, mill tailings (active sites), and remedial action wastes. The latter category includes mill tailings (inactive sites), surplus facilities, formerly utilized sites, and the Grand Junction Project. For each category, waste volume inventories and projections are given through the year 2000. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for present inventories of mill tailings

  8. Regional nuclear fuel cycle centers study project

    International Nuclear Information System (INIS)

    Bennett, L.; Catlin, R.G.; Meckoni, V.

    1977-01-01

    The concept of regional fuel cycle centers (RFCC) has attracted wide interest. The concept was endorsed by many countries in discussions at the General Conference of the International Atomic Energy Agency and at the General Assembly of the United Nations. Accordingly, in 1975, the IAEA initiated a detailed study of the RFCC concept. The Agency study has concentrated on what is referred to as the ''back-end'' of the fuel cycle because that is the portion which is currently problematic. The study covers transport, storage, processing and recycle activities starting from the time the spent fuel leaves the reactor storage pools and through all steps until the recycled fuel is in finished fuel elements and shipped to the reactor. A detailed evaluation of the specific features of large regional fuel cycle centers established on a multinational basis vis-a-vis smaller dispersed fuel cycle facilities set up on a national basis has been carried out. The methodology for assessment of alternative strategies for fuel storage, reprocessing, and recycling of plutonium has been developed, characteristic data on material flows and cost factors have been generated, and an analytic system has been developed to carry out such evaluations including appropriate sensitivity analysis. Studies in related areas on institutional and legal, organizational, environmental, materials control and other essential aspects have also been made. The material developed during the course of this Study would enable any group of interested Member States to examine and work out alternative strategies pertinent to their present and projected nuclear fuel cycle needs, as well as evolve institutional, legal and other appropriate frameworks or agreements for the establishment of fuel cycle centers on a multinational cooperative basis

  9. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    Pajunen, A.L.

    1998-01-01

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

  10. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    1999-01-01

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project

  11. Hanford environmental dose reconstruction project - an overview

    International Nuclear Information System (INIS)

    Shipler, D.B.; Napier, B.A.; Farris, W.T.

    1996-01-01

    The Hanford Environmental Dose Reconstruction Project was initiated because of public interest in the historical releases of radioactive materials from the Hanford Site, located in southcentral Washington State. By 1986, over 38,000 pages of environmental monitoring documentation from the early years of Hanford operations had been released. Special committees reviewing the documents recommended initiation of the Hanford Environmental Dose Reconstruction Project, which began in October 1987, and is conducted by Battelle, Pacific Northwest Laboratories. The technical approach taken was to reconstruct releases of radioactive materials based on facility operating information; develop and/or adapt transport, pathway, and dose models and computer codes; reconstruct environmental, meterological, and hydrological monitoring information; reconstruct demographic, agricultural, and lifestyle characteristics; apply statistical methods to all forms of uncertainty in the information, parameters, and models; and perform scientific investigation that were technically defensible. The geographic area for the study includes ∼2 x 10 5 km 2 (75,000 mi 2 ) in eastern Washington, western Idaho, and northeastern Oregon (essentially the Mid-columbia Basin of the Pacific Northwest). Three exposure pathways were considered: the atmosphere, the Columbia River, and ground water

  12. Concrete spent fuel storage casks dose rates

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Trontl, K.

    1998-01-01

    Our intention was to model a series of concrete storage casks based on TranStor system storage cask VSC-24, and calculate the dose rates at the surface of the casks as a function of extended burnup and a prolonged cooling time. All of the modeled casks have been filled with the original multi-assembly sealed basket. The thickness of the concrete shield has been varied. A series of dose rate calculations for different burnup and cooling time values have been performed. The results of the calculations show rather conservative original design of the VSC-24 system, considering only the dose rate values, and appropriate design considering heat rejection.(author)

  13. Hanford spent nuclear fuel project update

    Energy Technology Data Exchange (ETDEWEB)

    Williams, N.H.

    1997-08-19

    Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building.

  14. American fuel cell bus project : first analysis report.

    Science.gov (United States)

    2013-06-01

    This report summarizes the experience and early results from the American Fuel Cell Bus Project, a fuel cell electric bus demonstration : funded by the Federal Transit Administration (FTA) under the National Fuel Cell Bus Program. A team led by CALST...

  15. Spent nuclear fuel project technical databook

    International Nuclear Information System (INIS)

    Reilly, M.A.

    1998-01-01

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values

  16. Spent nuclear fuel project technical databook

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, M.A.

    1998-07-22

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values.

  17. Southern Nevada Alternative Fuels Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Hyde, Dan; Fast, Matthew

    2009-12-31

    The Southern Nevada Alternative Fuels Program is designed to demonstrate, in a day-to-day bus operation, the reliability and efficiency of a hydrogen bus operation under extreme conditions. By using ICE technology and utilizing a virtually emission free fuel, benefits to be derived include air quality enhancement and vehicle performance improvements from domestically produced, renewable energy sources. The project objective is to help both Ford and the City demonstrate and evaluate the performance characteristics of the E-450 H2ICE shuttle buses developed by Ford, which use a 6.8-liter supercharged Triton V-10 engine with a hydrogen storage system equivalent to 29 gallons of gasoline. The technology used during the demonstration project in the Ford buses is a modified internal combustion engine that allows the vehicles to run on 100% hydrogen fuel. Hydrogen gives a more thorough fuel burn which results in more power and responsiveness and less pollution. The resultant emissions from the tailpipe are 2010 Phase II compliant with NO after treatment. The City will lease two of these E-450 H2ICE buses from Ford for two years. The buses are outfitted with additional equipment used to gather information needed for the evaluation. Performance, reliability, safety, efficiency, and rider comments data will be collected. The method of data collection will be both electronically and manually. Emissions readings were not obtained during the project. The City planned to measure the vehicle exhaust with an emissions analyzer machine but discovered the bus emission levels were below the capability of their machine. Passenger comments were solicited on the survey cards. The majority of comments were favorable. The controllable issues encountered during this demonstration project were mainly due to the size of the hydrogen fuel tanks at the site and the amount of fuel that could be dispensed during a specified period of time. The uncontrollable issues encountered during this

  18. Canadian Fusion Fuels Technology Project activities report

    International Nuclear Information System (INIS)

    1985-01-01

    The Canadian Fusion Fuels Technology Project was formally established in 1982. The project is directed toward the further development of Canadian capabilities in five major areas: tritium technology, breeder technology, materials technology, equipment development and safety and the environment. The project is funded by three partners - Government of Canada (50%), Ontario Provincial Government (25%) and Ontario Hydro (25%). The fiscal year 1984/85 represents the third year of operation of the project. In 1984/85, 108 contracts were awarded totalling $4 million. Supplementary funding by subcontractors added approximately $1.9 million to the total project value. More than 200 people participated in the technical work involved in the project. Sixteen people were on attachment to foreign facilities for terms ranging from 1 month to 2.5 years. Five patents were applied for including a tritium discrimination monitor, a new radio-chemical tritium separation method, a new variation of fuel cleanup by gas chromatography, a passive tritium permeation system using bimetallic membranes, and a new breeder process using lithium salts dissolved in heavy water

  19. The Canadian Fusion Fuels Technology Project

    International Nuclear Information System (INIS)

    Dautovich, D.P.; Gierszewski, P.J.; Wong, K.Y.; Stasko, R.R.; Burnham, C.D.

    1987-04-01

    The Canadian Fusion Fuels Technology Project (CFFTP) is a national project whose aim is to develop capability in tritium and robotics technologies for application to international fusion development programs. Activities over the first five years have brought substantial interaction with the world's leading projects such as Tokamak Fusion Test Reactor (TFTR), the Joint European Torus (JET), and the Next European Torus (NET), Canadian R and D and engineering services, and hardware are in demand as these major projects prepare for tritium operation leading to the demonstration of energy breakeven around 1990. Global planning is underway for the next generation ignition experiment. It is anticipated this will provide increased opportunity for CFFTP and its contractors among industry, universities and governmental laboratories

  20. NAC-1 cask dose rate calculations for LWR spent fuel

    International Nuclear Information System (INIS)

    CARLSON, A.B.

    1999-01-01

    A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation

  1. Hanford Environmental Dose Reconstruction Project monthly report, August 1992

    International Nuclear Information System (INIS)

    McMakin, A.H.; Cannon, S.D.; Finch, S.M.

    1992-01-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation doses that individuals and populations could have received from nuclear operations at Hanford since 1944. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed from release to impact on humans (dose estimates): source terms; environmental transport; environmental monitoring data; demography; food consumption; and agriculture; and environmental pathway and dose estimates

  2. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    International Nuclear Information System (INIS)

    Norton, S.H.

    2010-01-01

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  3. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    NORTON SH

    2010-02-23

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  4. Fuels planning: science synthesis and integration; fact sheet: The Fuels Synthesis Project overview

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    The geographic focus of the "Fuels Planning: Science Synthesis and Integration" project #known as the Fuels Synthesis Project# is on the dry forests of the Western United States. Target audiences include fuels management specialists, resource specialists, National Environmental Policy Act #NEPA# planning team leaders, line officers in the USDA Forest Service...

  5. Fuel Retrieval Sub-Project (FRS) Stuck Fuel Station Performance Test Data Report

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2000-01-01

    This document provides the test data report for Stuck Fuel Station Performance Testing in support of the Fuel Retrieval Sub-Project. The stuck fuel station was designed to provide a means of cutting open a canister barrel to release fuel elements, etc

  6. Spent Nuclear Fuel (SNF) Project Product Specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  7. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  8. Overview of the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Shipler, D.B.; Napier, B.A.; Ikenberry, T.A.

    1992-04-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation doses that specific and representative individuals and populations may have received as a result of releases of radioactive materials from historical operations at the Hanford Site. These dose estimates would account for the uncertainties of information regarding facilities operations, environmental monitoring, demography, food consumption and lifestyles, and the variability of natural phenomena. Other objectives of the HEDR Project include: supporting the Hanford Thyroid Disease Study (HTDS), declassifying Hanford-generated information and making it available to the public, performing high-quality, credible science, and conducting the project in an open, public forum. The project is briefly described

  9. Radiation dose rates from commercial PWR and BWR spent fuel elements

    International Nuclear Information System (INIS)

    Willingham, C.E.

    1981-10-01

    Data on measurements of gamma dose rates from commercial reactor spent fuel were collected, and documented calculated gamma dose rates were reviewed. As part of this study, the gamma dose rate from spent fuel was estimated, using computational techniques similar to previous investigations into this problem. Comparison of the measured and calculated dose rates provided a recommended dose rate in air versus distance curve for PWR spent fuel

  10. An examination of fuel consumption trends in construction projects

    International Nuclear Information System (INIS)

    Peters, Valerie A.; Manley, Dawn K.

    2012-01-01

    Recent estimates of fuel consumption in construction projects are highly variable. Lack of standards for reporting at both the equipment and project levels make it difficult to quantify the magnitude of fuel consumption and the associated opportunities for efficiency improvements in construction projects. In this study, we examined clusters of Environmental Impact Reports for seemingly similar construction projects in California. We observed that construction projects are not characterized consistently by task or equipment. We found wide variations in estimates for fuel use in terms of tasks, equipment, and overall projects, which may be attributed in part to inconsistencies in methodology and parameter ranges. Our analysis suggests that standardizing fuel consumption reporting and estimation methodologies for construction projects would enable quantification of opportunities for efficiency improvements at both the equipment and project levels. With increasing emphasis on reducing fossil fuel consumption, it will be important to quantify opportunities to increase fuel efficiency, including across the construction sector. - Highlights: ► An analysis of construction projects reveals inconsistencies in fuel use estimates. ► Fuel consumption estimates for similar construction equipment can vary greatly. ► Standards would help to quantify efficiency opportunities in construction.

  11. Population dose from nuclear medicine studies (DOMNES). Contribution of Project DOMNES to Dose Datemed2

    International Nuclear Information System (INIS)

    Ramirez, M. L.; Ruiz, A.; Ferrer, N.; Alonso Farto, J. C.; Alvarez, C.; Rodriguez, M.

    2013-01-01

    The DOMNES Project is created in 2001 to carry out a survey on nuclear medicine procedures used in the Spanish health centers, their frequency and the doses given to patients. In addition, it reports information to Dose Data Project Med 2, focusing on radiology exams. (Author)

  12. Spent nuclear fuel project quality assurance program plan

    International Nuclear Information System (INIS)

    Lacey, R.E.

    1997-01-01

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures

  13. Canadian Fusion Fuels Technology Project annual report 93/94

    International Nuclear Information System (INIS)

    1994-01-01

    The Canadian Fusion Fuels Technology Project exists to develop fusion technologies and apply them worldwide in today's advanced fusion projects and to apply these technologies in fusion and tritium research facilities. CFFTP concentrates on developing capability in fusion fuel cycle systems, in tritium handling technologies and in remote handling. This is an annual report for CFFTP and as such also includes a financial report

  14. Work plan for the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    1989-12-01

    The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation doses that populations could have received from nuclear operations at the Hanford Site since 1944, with descriptions of uncertainties inherent in such estimates. The secondary objective is to make project records--information that HEDR staff members used to estimate radiation doses--available to the public. Preliminary dose estimates for a limited geographic area and time period, certain radionuclides, and certain populations are planned to be available in 1990; complete results are planned to be reported in 1993. Project reports and references used in the reports are available to the public in the DOE Public Reading Room in Richland, Washington. Project progress is documented in monthly reports, which are also available to the public in the DOE Public Reading Room.

  15. Spent nuclear fuel project design basis capacity study

    International Nuclear Information System (INIS)

    Cleveland, K.J.

    1998-01-01

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. A base case, reflecting the Fiscal Year 1998 process configuration, is evaluated. Parametric evaluations are also considered, investigating the impact of higher fuel retrieval system productivity and reduced shift operations at the canister storage building on total project duration

  16. Spent Nuclear Fuel project interface control plan

    International Nuclear Information System (INIS)

    Reilly, M.A.

    1995-01-01

    This implementation process philosophy is in keeping with the ongoing reengineering of the WHC Controlled Manuals to achieve interface control within the SNF Project. This plan applies to all SNF Project sub-project to sub-project, and sub-project to exteranl (both on and off the Hanford Site) interfaces

  17. Spent nuclear fuel project design basis capacity study

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, K.J.

    1996-09-09

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. Alternative configurations, sub-system cycle times, and operating scenarios were tested to identify their impact on total project duration and equipment requirements.

  18. Integration of models for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Napier, B.A.

    1991-01-01

    The objective of the Hanford Environmental Dose Reconstruction Project is to estimate the radiation dose that individuals could have received as a result of emissions from nuclear operations at Hanford since 1944. The objective of phase 1 of the project was to demonstrate through calculations that adequate models and support data exist or could be developed to allow realistic estimations of doses to individuals from releases of radionuclides to the environment that occurred as long as 45 years ago. Much of the data used in phase 1 was preliminary; therefore, the doses calculated must be considered preliminary approximations. This paper describes the integration of various models that was implemented for initial computer calculations. Models were required for estimating the quantity of radioactive material released, for evaluating its transport through the environment, for estimating human exposure, and for evaluating resultant doses

  19. Updating Dosimetry for Emergency Response Dose Projections.

    Science.gov (United States)

    DeCair, Sara

    2016-02-01

    In 2013, the U.S. Environmental Protection Agency (EPA) proposed an update to the 1992 Protective Action Guides (PAG) Manual. The PAG Manual provides guidance to state and local officials planning for radiological emergencies. EPA requested public comment on the proposed revisions, while making them available for interim use by officials faced with an emergency situation. Developed with interagency partners, EPA's proposal incorporates newer dosimetric methods, identifies tools and guidelines developed since the current document was issued, and extends the scope of the PAGs to all significant radiological incidents, including radiological dispersal devices or improvised nuclear devices. In order to best serve the emergency management community, scientific policy direction had to be set on how to use International Commission on Radiological Protection Publication 60 age groups in dose assessment when implementing emergency guidelines. Certain guidelines that lend themselves to different PAGs for different subpopulations are the PAGs for potassium iodide (KI), food, and water. These guidelines provide age-specific recommendations because of the radiosensitivity of the thyroid and young children with respect to ingestion and inhalation doses in particular. Taking protective actions like using KI, avoiding certain foods or using alternative sources of drinking water can be relatively simple to implement by the parents of young children. Clear public messages can convey which age groups should take which action, unlike how an evacuation or relocation order should apply to entire households or neighborhoods. New in the PAG Manual is planning guidance for the late phase of an incident, after the situation is stabilized and efforts turn toward recovery. Because the late phase can take years to complete, decision makers are faced with managing public exposures in areas not fully remediated. The proposal includes quick-reference operational guidelines to inform re-entry to

  20. Prototypical consolidation demonstration project - Final fuel recommendation report

    International Nuclear Information System (INIS)

    Piscitella, R.R.; Paskey, W.R.

    1987-01-01

    The Prototypical Consolidation Demonstration (PCD) Project will, in its final phase, conduct a demonstration of the equipment's ability to consolidate actual spent commercial fuel. Since budget and schedule limitations do not allow this demonstration to include all types of fuel assemblies, a selection process was utilized to identify the fuel types that would represent predominate fuel inventories and that would demonstrate the equipment's abilities. The Pressurized Water Reactor (PWR) fuel assemblies that were suggested for use in the PCD Project Hot Demonstration were Babcock and Wilcox (B and W) 15 x 15's, and Westinghouse (WE) 15 x 15's. The Boiling Water Reactor (BWR) fuel suggested was the General Electric (GE) 8 x 8

  1. Spent fuel stability under repository conditions - final report of the european project

    International Nuclear Information System (INIS)

    Poinssot, Ch.; Ferry, C.; Kelm, M.; Cavedon, J.M.; Corbel, C.; Jegou, Ch.; Lovera, P.; Miserque, F.; Poulesquen, A.; Grambow, B.; Andriambololona, Z.; Martinez-Esparza, A.; Kelm, M.; Loida, A.; Rondinella, V.; Wegen, D.; Spahiu, K.; Johnson, L.; Cachoir, Ch.; Lemmens, K.; Quinones, J.; Bruno, J.; Christensen, H.; Grambow, B.; Pablo, J. de

    2005-01-01

    This report is the final report of the European Project 'Spent Fuel Stability under Repository Conditions' (FIKW-CT-2001-00192 SFS) funded by the European Commission from Nov.2000 to Oct.2004. Gathering the work performed by 13 partners from 6 countries, it aims to specifically focus on the spent nuclear fuel long term alteration in deep repository and the subsequent radionuclides release rate as a function of time. This report synthesised the wide experimental work performed within this project and enlightens the major outcomes, which can be summarised as follow: - A new model for defining the Instant Release Fraction was developed in order to consider the potential fuel evolution before the water penetrates the canister. Quantitative assessment has been produced and shows a significant contribution to the long term dose; - Based on new experimental data, kinetic radiolytic scheme have been upgraded and are used to determine the amount of oxidants produced at the fuel/water interface; - The existence of a dose threshold below which the water radiolysis does not influence the fuel alteration has been demonstrated and occurs between 3.5 and 33 MBq.g UO21. Above the threshold, the fuel alteration rates is directly related to the dose rate. - Hydrogen was experimentally demonstrated to be an efficient oxidants scavenger preventing therefore the fuel oxidation. Molecular mechanism still need to be understood. - Finally, a new Matrix Alteration Model integrating most of the SFS results (apart of the hydrogen effect) has been developed and used to assess the fuel long tern stability in representative conditions of deep repository in salt, clay-rock and granite. The breadth of the results and the significance of the conclusions testify of the success of the collaboration within the project. (authors)

  2. Projection of US LWR spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.F.; Cole, B.M.; Purcell, W.L.; Rau, R.G.

    1982-11-01

    The spent fuel storage requirements projection is based on data supplied for each operating or planned nuclear power power plant by the operting utilities. The data supplied by the utilities encompassed details of plant operating history, past records of fuel discharges, current inventories in reactor spent fuel storage pools, and projections of future discharge patterns. Data on storage capacity of storage pools and on characterization of the discharged fuel are also included. The data supplied by the utilities, plus additional data from other appropriate sources, are maintained on a computerized data base by Pacific Northwest Laboratory. The spent fuel requirements projection was based on utility data updated and verified as of December 31, 1981

  3. Studsvik`s fuel R and D projects

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M [Studsvik Nuclearr AB, Nykoping (Sweden)

    1997-08-01

    The report reviews some recently performed, ongoing and planned fuel R and D projects, executed by Studsvik Nuclear AB, a subsidiary of Studsvik AB. Data from these projects are used as experimental support for fuel modelling at high burnup. Much of Studsvik Nuclear`s R and D work has been concentrated on fuel testing, which can be made in the R2 test reactor with high precision under realistic water reactor conditions. This type of work started in the early 1960s. The fuel testing projects executed at Studsvik have been organized under three different types of sponsorship: International (multilateral) fuel projects: jointly sponsored internationally on a world-wide basis, with project information remaining restricted to the project participants throughout the project`s duration and for some pre-determined time after project completion; Bilateral fuel projects: sponsored by one single organization, or a few co-operating organizations, with project information remaining restricted to the sponsor, sometimes published later; in-house R and D work: sponsored by Studsvik Nuclear. The fuel testing activities can be divided into a number of well-defined steps as follows: Base irradiation, performed in a power reactor, or in Studsvik`s R2 test reactor; power ramping and/or other in-pile measurements, performed in Studsvik`s R2 test reactor. Non-destructive testing between different phases of an experiment, performed in Studsvik`s R2 reactor pool, or in Studsvik`s Hot Cell Laboratory; destructive post-irradiation examinations, performed in Studsvik`s Hot Cell Laboratory, or in the sponsor`s hot cell laboratory. 47 refs, 2 tabs.

  4. Phase 1 of the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    1991-08-01

    The work described in this report was prompted by the public's concern about potential effect from the radioactive materials released from the Hanford Site. The Hanford Environmental Dose Reconstruction (HEDR) Project was established to estimate radiation dose the public might have received from the Hanford Site since 1944, when facilities began operating. Phase 1 of the HEDR Project is a ''pilot'' or ''demonstration'' phase. The objectives of this initial phase were to determine whether enough historical information could be found or reconstructed to be used for dose estimation and develop and test conceptual and computational models for calculating credible dose estimates. Preliminary estimates of radiation doses were produced in Phase 1 because they are needed to achieve these objectives. The reader is cautioned that the dose estimates provided in this and other Phase 1 HEDR reports are preliminary. As the HEDR Project continues, the dose estimates will change for at least three reasons: more complete input information for models will be developed; the models themselves will be refined; and the size and shape of the geographic study area will change. This is one of three draft reports that summarize the first phase of the four-phased HEDR Project. This, the Summary Report, is directed to readers who want a general understanding of the Phase 1 work and preliminary dose estimates. The two other reports -- the Air Pathway Report and the Columbia River Pathway Report -- are for readers who understand the radiation dose assessment process and want to see more technical detail. Detailed descriptions of the dose reconstruction process are available in more than 20 supporting reports listed in Appendix A. 32 refs., 46 figs

  5. Progress on the Hanford K basins spent nuclear fuel project

    International Nuclear Information System (INIS)

    Culley, G.E.; Fulton, J.C.; Gerber, E.W.

    1996-01-01

    This paper highlights progress made during the last year toward removing the Department of Energy's (DOE) approximately, 2,100 metric tons of metallic spent nuclear fuel from the two outdated K Basins at the Hanford Site and placing it in safe, economical interim dry storage. In the past year, the Spent Nuclear Fuel (SNF) Project has engaged in an evolutionary process involving the customer, regulatory bodies, and the public that has resulted in a quicker, cheaper, and safer strategy for accomplishing that goal. Development and implementation of the Integrated Process Strategy for K Basins Fuel is as much a case study of modern project and business management within the regulatory system as it is a technical achievement. A year ago, the SNF Project developed the K Basins Path Forward that, beginning in December 1998, would move the spent nuclear fuel currently stored in the K Basins to a new Staging and Storage Facility by December 2000. The second stage of this $960 million two-stage plan would complete the project by conditioning the metallic fuel and placing it in interim dry storage by 2006. In accepting this plan, the DOE established goals that the fuel removal schedule be accelerated by a year, that fuel conditioning be closely coupled with fuel removal, and that the cost be reduced by at least $300 million. The SNF Project conducted coordinated engineering and technology studies over a three-month period that established the technical framework needed to design and construct facilities, and implement processes compatible with these goals. The result was the Integrated Process Strategy for K Basins Fuel. This strategy accomplishes the goals set forth by the DOE by beginning fuel removal a year earlier in December 1997, completing it by December 1999, beginning conditioning within six months of starting fuel removal, and accomplishes it for $340 million less than the previous Path Forward plan

  6. Solar Hydrogen Fuel Cell Projects at Brooklyn Tech

    Science.gov (United States)

    Fedotov, Alex; Farah, Shadia; Farley, Daithi; Ghani, Naureen; Kuo, Emmy; Aponte, Cecielo; Abrescia, Leo; Kwan, Laiyee; Khan, Ussamah; Khizner, Felix; Yam, Anthony; Sakeeb, Khan; Grey, Daniel; Anika, Zarin; Issa, Fouad; Boussayoud, Chayama; Abdeldayem, Mahmoud; Zhang, Alvin; Chen, Kelin; Chan, Kameron Chuen; Roytman, Viktor; Yee, Michael

    2010-01-01

    This article describes the projects on solar hydrogen powered vehicles using water as fuel conducted by teams at Brooklyn Technical High School. Their investigations into the pure and applied chemical thermodynamics of hydrogen fuel cells and bio-inspired devices have been consolidated in a new and emerging sub-discipline that they define as solar…

  7. Nuclear Fuel Assembly Assessment Project and Image Categorization

    Energy Technology Data Exchange (ETDEWEB)

    Lindsey, C.S.; Lindblad, T.; Waldemark, K. [Royal Inst. of Tech., Stockholm (Sweden); Hildingsson, Lars [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    1998-07-01

    A project has been underway to add digital imaging and processing to the inspection of nuclear fuel by the International Atomic Energy Agency. The ultimate goals are to provide the inspector not only with the advantages of Ccd imaging, such as high sensitivity and digital image enhancements, but also with an intelligent agent that can analyze the images and provide useful information about the fuel assemblies in real time. The project is still in the early stages and several interesting sub-projects have been inspired. Here we give first a review of the work on the fuel assembly image analysis and then give a brief status report on one of these sub-projects that concerns automatic categorization of fuel assembly images. The technique could be of benefit to the general challenge of image categorization

  8. Spent nuclear fuel project integrated schedule plan

    International Nuclear Information System (INIS)

    Squires, K.G.

    1995-01-01

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel

  9. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

  10. Low NOx Fuel Flexible Combustor Integration Project Overview

    Science.gov (United States)

    Walton, Joanne C.; Chang, Clarence T.; Lee, Chi-Ming; Kramer, Stephen

    2015-01-01

    The Integrated Technology Demonstration (ITD) 40A Low NOx Fuel Flexible Combustor Integration development is being conducted as part of the NASA Environmentally Responsible Aviation (ERA) Project. Phase 2 of this effort began in 2012 and will end in 2015. This document describes the ERA goals, how the fuel flexible combustor integration development fulfills the ERA combustor goals, and outlines the work to be conducted during project execution.

  11. Canadian Fusion Fuels Technology Project annual report 93/94

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The Canadian Fusion Fuels Technology Project exists to develop fusion technologies and apply them worldwide in today`s advanced fusion projects and to apply these technologies in fusion and tritium research facilities. CFFTP concentrates on developing capability in fusion fuel cycle systems, in tritium handling technologies and in remote handling. This is an annual report for CFFTP and as such also includes a financial report.

  12. Individual and collective doses associated with the transport of irradiated magnox fuel within the UK

    International Nuclear Information System (INIS)

    Macdonald, H.F.; Mairs, J.H.

    1978-12-01

    A method is described of evaluating the individual and collective doses arising during the transport of irradiated fuel from a system of nuclear power stations to a central reprocessing plant. The doses associated with irradiated Magnox fuel movements in the UK are estimated and compared with those resulting from other phases of the nuclear fuel cycle. In addition, the individual and collective doses implied by the accidental activity release limits contained within the 1973 IAEA Tranport Regulations are discussed. (author)

  13. Population doses in Spain. Contribution of the project dopoes a dose Datamed 2

    International Nuclear Information System (INIS)

    Ruiz Cruces, R.; Canete Hidalgo, S.; Perez Martinez, M.; Pola, A.; Moreno, S.; Rodriguez, M.; Alvarez, C.; Gil, M.

    2013-01-01

    Frequency and effective dose values are of the order of the reported in the publication Radiation Protection 154 by neighbouring countries. Spain participated actively in the project DDM2 by sending all the required information and this has served test to ensure the correct development of the DOPOES project, which is ongoing. (Author)

  14. Comparison between the chest dose and the neck dose of workers with protective aprons at PNC plutonium fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, Norio; Momose, Takumaro; Shinohara, Kunihiko [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1996-06-01

    The dose equivalents recorded by a chest dosemeter under the protective apron and a neck dosemeter above the apron, worn by workers in the fabrication process of MOX fuels at PNC Tokai works, are compared. The ratio of the chest and neck dose equivalent is from 3 to 4. The effective dose equivalent calculated from a weighted combination of the dosemeter readings is about 2 times of the dose under protective aprons. (author)

  15. Comparison between the chest dose and the neck dose of workers with protective aprons at PNC plutonium fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Momose, Takumaro; Shinohara, Kunihiko

    1996-01-01

    The dose equivalents recorded by a chest dosemeter under the protective apron and a neck dosemeter above the apron, worn by workers in the fabrication process of MOX fuels at PNC Tokai works, are compared. The ratio of the chest and neck dose equivalent is from 3 to 4. The effective dose equivalent calculated from a weighted combination of the dosemeter readings is about 2 times of the dose under protective aprons. (author)

  16. The continual fuel management modification in Qinshan project II

    International Nuclear Information System (INIS)

    Ye Guodong; Pan Zefei; Zhang Xingtian

    2010-01-01

    The fuel management strategy is the basis of the nuclear power plants. The performance of the fuel management strategy affects the plants' safety and economy indicators directly. The paper summarizes all the modifications on the fuel management work in Qinshan Project II since the plant was established. It includes the surveillance system of physics tests, fetching in high performance fuel assemblies, reloading pattern optimization, and the modifications of the final safety analysis report. At the same time, it evaluates the benefit of the modifications in the few years. The experience in this paper is much helpful and could be implemented on the same type plants. (authors)

  17. DIMETHYL ETHER (DME)-FUELED SHUTTLE BUS DEMONSTRATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Elana M. Chapman; Shirish Bhide; Jennifer Stefanik; Andre L. Boehman; David Klinikowski

    2003-04-01

    The objectives of this research and demonstration program are to convert a campus shuttle bus to operation on dimethyl ether, a potential ultra-clean alternative diesel fuel. To accomplish this objective, this project includes laboratory evaluation of a fuel conversion strategy, as well as, field demonstration of the DME-fueled shuttle bus. Since DME is a fuel with no lubricity (i.e., it does not possess the lubricating quality of diesel fuel), conventional fuel delivery and fuel injection systems are not compatible with dimethyl ether. Therefore, to operate a diesel engine on DME one must develop a fuel-tolerant injection system, or find a way to provide the necessary lubricity to the DME. In this project, they have chosen the latter strategy in order to achieve the objective with minimal need to modify the engine. The strategy is to blend DME with diesel fuel, to obtain the necessary lubricity to protect the fuel injection system and to achieve low emissions. The laboratory studies have included work with a Navistar V-8 turbodiesel engine, demonstration of engine operation on DME-diesel blends and instrumentation for evaluating fuel properties. The field studies have involved performance, efficiency and emissions measurements with the Champion Motorcoach ''Defender'' shuttle bus which will be converted to DME-fueling. The results include baseline emissions, performance and combustion measurements on the Navistar engine for operation on a federal low sulfur diesel fuel (300 ppm S). Most recently, they have completed engine combustion studies on DME-diesel blends up to 30 wt% DME addition.

  18. Communication tools for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Blazek, Mary Lou; Power, Max S.

    1992-01-01

    From 1944 to 1989, the U.S. Department of Energy produced plutonium at the Hanford Site in southeast Washington State. In the early days of operation, large amounts of radioactive materials were released to the environment. Documents about the releases were made public in 1986. The Hanford Environmental Dose Reconstruction Project began in 1987. The Project will determine how much radioactive material was released, how that material may have exposed people, and what radiation doses people may have received. The Project will be complete in 1995. The federal government pays for the work. The scientific work on the study is done by Battelle's Pacific Northwest Laboratory. Public credibility and valid science are equally important to those directing the dose reconstruction work. A number of tools are used to inform the public and encourage public participation. These tools are examined in this paper. (author)

  19. Final design review report for K basin dose reduction project

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1996-01-01

    The strategy for reducing radiation dose originating from radionuclides absorbed in the K East Basin concrete is to raise the pool water level to provide additional shielding. This report documents a final design review for cleaning/coating basin walls and modifying other basin components where appropriate. The conclusion of this review was that the documents developed constitute an acceptable design for the Dose Reduction Project

  20. Decision management for the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Roberds, W.J.; Haerer, H.A. [Golder Associates, Inc., Redmond, WA (United States); Winterfeldt, D.V. [Decision Insights, Laguna Beach, CA (United States)

    1992-04-01

    The Hanford Environmental Dose Reconstruction (HEDR) Project is in the process of developing estimates for the radiation doses that individuals and population groups may have received as a result of past activities at the Hanford Reservation in Eastern Washington. A formal decision-aiding methodology has been developed to assist the HEDR Project in making significant and defensible decisions regarding how this study will be conducted. These decisions relate primarily to policy (e.g., the appropriate level of public participation in the study) and specific technical aspects (e.g., the appropriate domain and depth of the study), and may have significant consequences with respect to technical results, costs, and public acceptability.

  1. Decision management for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Roberds, W.J.; Haerer, H.A.; Winterfeldt, D.V.

    1992-04-01

    The Hanford Environmental Dose Reconstruction (HEDR) Project is in the process of developing estimates for the radiation doses that individuals and population groups may have received as a result of past activities at the Hanford Reservation in Eastern Washington. A formal decision-aiding methodology has been developed to assist the HEDR Project in making significant and defensible decisions regarding how this study will be conducted. These decisions relate primarily to policy (e.g., the appropriate level of public participation in the study) and specific technical aspects (e.g., the appropriate domain and depth of the study), and may have significant consequences with respect to technical results, costs, and public acceptability

  2. Overview of the spent nuclear fuel project at Hanford

    International Nuclear Information System (INIS)

    Daily, J.L.

    1995-02-01

    The Spent Nuclear Fuel Project's mission at Hanford is to open-quotes Provide safe, economic and environmentally sound management of Hanford spent nuclear fuel in a manner which stages it to final disposition.close quotes The inventory of spent nuclear fuel (SNF) at the Hanford Site covers a wide variety of fuel types (production reactor to space reactor) in many facilities (reactor fuel basins to hot cells) at locations all over the Site. The 2,129 metric tons of Hanford SNF represents about 80% of the total US Department of Energy (DOE) inventory. About 98.5% of the Hanford SNF is 2,100 metric tons of metallic uranium production reactor fuel currently stored in the 1950s vintage K Basins in the 100 Area. This fuel has been slowly corroding, generating sludge and contaminating the basin water. This condition, coupled with aging facilities with seismic vulnerabilities, has been identified by several groups, including stakeholders, as being one of the most urgent safety and environmental concerns at the Hanford Site. As a direct result of these concerns, the Spent Nuclear Fuel Project was recently formed to address spent fuel issues at Hanford. The Project has developed the K Basins Path Forward to remove fuel from the basins and place it in dry interim storage. Alternatives that addressed the requirements were developed and analyzed. The result is a two-phased approach allowing the early removal of fuel from the K Basins followed by its stabilization and interim storage consistent with the national program

  3. Fuel Cell Car Design Project for Freshman Engineering Courses

    Science.gov (United States)

    Duke, Steve R.; Davis, Virginia A.

    2014-01-01

    In the Samuel Ginn College of Engineering at Auburn University, we have integrated a semester long design project based on a toy fuel cell car into our freshman "Introduction to Chemical Engineering Class." The project provides the students a basic foundation in chemical reactions, energy, and dimensional analysis that facilitates…

  4. Spatial fuel data products of the LANDFIRE Project

    Science.gov (United States)

    Matt Reeves; Kevin C. Ryan; Matthew G. Rollins; Thomas G. Thompson

    2009-01-01

    The Landscape Fire and Resource Management Planning Tools (LANDFIRE) Project is mapping wildland fuels, vegetation, and fire regime characteristics across the United States. The LANDFIRE project is unique because of its national scope, creating an integrated product suite at 30-m spatial resolution and complete spatial coverage of all lands within the 50...

  5. Fuel price and supply projections, 1980 to 2000

    International Nuclear Information System (INIS)

    1980-06-01

    In 1978, over 95% of California's energy was derived from conventional fuels - oil, natural gas, coal, and uranium. Approximately one-third of these conventional fuels was produced within the state, the remaining two-thirds coming from other states and foreign countries. Dependence on these fuels is not likely to diminish rapidly in the near future, therefore the factors that contribute to the future supplies and prices of these fuels will have a major influence on the state's energy future. This report serves as a basis for Commission analysis and is also intended as a tool to be used by others who must make decisions involving the future cost and availability of fuels. This report documents the staff's projections on future supply, price, and availability of these fuels and presents information on historical fuel use and price for background and perspective. Analyses of commercially developable derived fuels and of recent Federal statutory restrictions on the use of oil and gas are also presented. These analyses include economic, logistic, environmental, geologic, and social and institutional considerations. This report does not focus on the costs included in fuel production and preparation; nor does the report go into detail on the transportation, disposal, and downstream costs of the various fuels

  6. Hydrogen fueling stations in Japan hydrogen and fuel cell demonstration project

    International Nuclear Information System (INIS)

    Koseki, K.; Tomuro, J.; Sato, H.; Maruyama, S.

    2004-01-01

    A new national demonstration project of fuel cell vehicles, which is called Japan Hydrogen and Fuel Cell Demonstration Project (JHFC Project), has started in FY2002 on a four-year plan. In this new project, ten hydrogen fueling stations have been constructed in Tokyo and Kanagawa area in FY2002-2003. The ten stations adopt the following different types of fuel and fueling methods: LPG reforming, methanol reforming, naphtha reforming, desulfurized-gasoline reforming, kerosene reforming, natural gas reforming, water electrolysis, liquid hydrogen, by-product hydrogen, and commercially available cylinder hydrogen. Approximately fifty fuel cell passenger cars and a fuel cell bus are running on public roads using these stations. In addition, two hydrogen stations will be constructed in FY2004 in Aichi prefecture where The 2005 World Exposition (EXPO 2005) will be held. The stations will service eight fuel cell buses used as pick-up buses for visitors. We, Engineering Advancement Association of Japan (ENAA), are commissioned to construct and operate a total of twelve stations by Ministry of Economy Trade and Industry (METI). We are executing to demonstrate or identify the energy-saving effect, reduction of the environmental footprint, and issues for facilitating the acceptance of hydrogen stations on the basis of the data obtained from the operation of the stations. (author)

  7. Spent Nuclear Fuel Project Configuration Management Plan

    International Nuclear Information System (INIS)

    Reilly, M.A.

    1995-01-01

    This document is a rewrite of the draft ''C'' that was agreed to ''in principle'' by SNF Project level 2 managers on EDT 609835, dated March 1995 (not released). The implementation process philosphy was changed in keeping with the ongoing reengineering of the WHC Controlled Manuals to achieve configuration management within the SNF Project

  8. Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

    2014-02-01

    This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

  9. Database for the OECD-IAEA Paks Fuel Project

    International Nuclear Information System (INIS)

    Szabo, Emese; Hozer, Zoltan; Gyori, Csaba; Hegyi, Gyoergy

    2010-01-01

    On 10 April 2003 severe damage of fuel assemblies took place during an incident at Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a special tank below the water level of the spent fuel storage pool in order to remove crud buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods were being cooled by circulation of storage pool water. The first sign of fuel failure was the detection of some fission gases released from the cleaning tank during that evening. The cleaning tank cover locks were released after midnight and this operation was followed by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel assemblies were severely damaged. The first evaluation of the event showed that the severe fuel damage happened due to inadequate coolant circulation within the cleaning tank. The damaged fuel assemblies will be removed from the cleaning tank in 2005 and will be stored in special canisters in the spent fuel storage pool of the Paks NPP. Following several discussions between expert from different countries and international organisations the OECD-IAEA Paks Fuel Project was proposed. The project is envisaged in two phases. - Phase 1 is to cover organization of visual inspection of material, preparation of database, performance of analyses and preparatory work for fuel examination. - Phase 2 is to cover the fuel transport and the hot cell examination. The first meeting of the project was held in Budapest on 30-31 January 2006. Phase 1 of the Paks Fuel Project will focus on the numerical simulation of the most important aspects of the incident. This activity will help in the reconstruction of the accidental scenario. The first step of Phase 1 was the collection of a database necessary for the code calculations. The main objective of database collection was to provide input data for calculations. For this reason the collection was focused on such data that are

  10. TMI defueling project fuel debris removal system

    International Nuclear Information System (INIS)

    Burdge, B.

    1992-01-01

    The three mile Island Unit 2 (TMI-2) pressurized water reactor loss-of-coolant accident on March 28, 1979, presented the nuclear community with many challenging remediation problems; most importantly, the removal of the fission products within the reactor containment vessel. To meet this removal problem, an air-lift system (ALS) can be used to employ compressed air to produce the motive force for transporting debris. Debris is separated from the transport stream by gravity separation. The entire method does not rely on any moving parts. Full-scale testing of the ALS at the Idaho National Engineering Laboratory (INEL) has demonstrated the capability of transporting fuel debris from beneath the LCSA into a standard fuel debris bucket at a minimum rate of 230 kg/min

  11. Constraints of fossil fuels depletion on global warming projections

    Energy Technology Data Exchange (ETDEWEB)

    Chiari, Luca, E-mail: chiari@science.unitn.it [Department of Physics, University of Trento, Via Sommarive 14, 38123 Povo (Italy); Zecca, Antonio, E-mail: zecca@science.unitn.it [Department of Physics, University of Trento, Via Sommarive 14, 38123 Povo (Italy)

    2011-09-15

    A scientific debate is in progress about the intersection of climate change with the new field of fossil fuels depletion geology. Here, new projections of atmospheric CO{sub 2} concentration and global-mean temperature change are presented, should fossil fuels be exploited at a rate limited by geological availability only. The present work starts from the projections of fossil energy use, as obtained from ten independent sources. From such projections an upper bound, a lower bound and an ensemble mean profile for fossil CO{sub 2} emissions until 2200 are derived. Using the coupled gas-cycle/climate model MAGICC, the corresponding climatic projections out to 2200 are obtained. We find that CO{sub 2} concentration might increase up to about 480 ppm (445-540 ppm), while the global-mean temperature increase w.r.t. 2000 might reach 1.2 deg. C (0.9-1.6 deg. C). However, future improvements of fossil fuels recovery and discoveries of new resources might lead to higher emissions; hence our climatic projections are likely to be underestimated. In the absence of actions of emissions reduction, a level of dangerous anthropogenic interference with the climate system might be already experienced toward the middle of the 21st century, despite the constraints imposed by the exhaustion of fossil fuels. - Highlights: > CO{sub 2} and global temperature are projected under fossil fuels exhaustion scenarios. > Temperature is projected to reach a minimum of 2 deg. C above pre-industrial. > Temperature projections are possibly lower than the IPCC ones. > Fossil fuels exhaustion will not avoid dangerous global warming.

  12. Constraints of fossil fuels depletion on global warming projections

    International Nuclear Information System (INIS)

    Chiari, Luca; Zecca, Antonio

    2011-01-01

    A scientific debate is in progress about the intersection of climate change with the new field of fossil fuels depletion geology. Here, new projections of atmospheric CO 2 concentration and global-mean temperature change are presented, should fossil fuels be exploited at a rate limited by geological availability only. The present work starts from the projections of fossil energy use, as obtained from ten independent sources. From such projections an upper bound, a lower bound and an ensemble mean profile for fossil CO 2 emissions until 2200 are derived. Using the coupled gas-cycle/climate model MAGICC, the corresponding climatic projections out to 2200 are obtained. We find that CO 2 concentration might increase up to about 480 ppm (445-540 ppm), while the global-mean temperature increase w.r.t. 2000 might reach 1.2 deg. C (0.9-1.6 deg. C). However, future improvements of fossil fuels recovery and discoveries of new resources might lead to higher emissions; hence our climatic projections are likely to be underestimated. In the absence of actions of emissions reduction, a level of dangerous anthropogenic interference with the climate system might be already experienced toward the middle of the 21st century, despite the constraints imposed by the exhaustion of fossil fuels. - Highlights: → CO 2 and global temperature are projected under fossil fuels exhaustion scenarios. → Temperature is projected to reach a minimum of 2 deg. C above pre-industrial. → Temperature projections are possibly lower than the IPCC ones. → Fossil fuels exhaustion will not avoid dangerous global warming.

  13. Hanford Spent Nuclear Fuel Project recommended path forward

    International Nuclear Information System (INIS)

    Fulton, J.C.

    1994-10-01

    The Spent Nuclear Fuel Project (the Project), in conjunction with the U.S. Department of Energy-commissioned Independent Technical Assessment (ITA) team, has developed engineered alternatives for expedited removal of spent nuclear fuel, including sludge, from the K Basins at Hanford. These alternatives, along with a foreign processing alternative offered by British Nuclear Fuels Limited (BNFL), were extensively reviewed and evaluated. Based on these evaluations, a Westinghouse Hanford Company (WHC) Recommended Path Forward for K Basins spent nuclear fuel has been developed and is presented in Volume I of this document. The recommendation constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. The overall processing and storage scheme is based on the ITA team's proposed passivation and vault storage process. A dual purpose staging and vault storage facility provides an innovative feature which allows accelerated removal of fuel and sludge from the basins and minimizes programmatic risks beyond any of the originally proposed alternatives. The projects fit within a regulatory and National Environmental Policy Act (NEPA) overlay which mandates a two-phased approach to construction and operation of the needed facilities. The two-phase strategy packages and moves K Basins fuel and sludge to a newly constructed Staging and Storage Facility by the year 2000 where it is staged for processing. When an adjoining facility is constructed, the fuel is cycled through a stabilization process and returned to the Staging and Storage Facility for dry interim (40-year) storage. The estimated total expenditure for this Recommended Path Forward, including necessary new construction, operations, and deactivation of Project facilities through 2012, is approximately $1,150 million (unescalated)

  14. Hydrogen fueling demonstration projects using compact PSA purification

    International Nuclear Information System (INIS)

    Ng, E.; Smith, T.

    2004-01-01

    'Full text:' Hydrogen fueling demonstration projects are critical to the success of hydrogen as an automotive fuel by building public awareness and demonstrating the technology required to produce, store, and dispense hydrogen. Over 75 of these demonstration projects have been undertaken or are in the planning stages world-wide, sponsored by both the public and private sectors. Each of these projects represents a unique combination of sponsors, participants, geographic location, and hydrogen production pathway. QuestAir Technologies Inc., as the industry leader in compact pressure swing adsorption equipment for purifying hydrogen, has participated in four hydrogen fueling demonstration projects with a variety of partners and in North America and Japan. QuestAir's experiences as a participant in the planning, construction, and commissioning of these demonstration projects will be presented in this paper. The unique challenges of each project and the critical success factors that must to be considered for successful deployment of high-profile, international, and multi-vendor collaborations will also be discussed. The paper will also provide insights on the requirements for hydrogen fueling demonstration projects in the future. (author)

  15. Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

    International Nuclear Information System (INIS)

    CLEVELAND, K.J.

    2000-01-01

    This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage

  16. American Fuel Cell Bus Project Evaluation. Second Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Post, Matthew [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-09-01

    This report presents results of the American Fuel Cell Bus (AFCB) Project, a demonstration of fuel cell electric buses operating in the Coachella Valley area of California. The prototype AFCB was developed as part of the Federal Transit Administration's (FTA's) National Fuel Cell Bus Program. Through the non-profit consortia CALSTART, a team led by SunLine Transit Agency and BAE Systems developed a new fuel cell electric bus for demonstration. SunLine added two more AFCBs to its fleet in 2014 and another in 2015. FTA and the AFCB project team are collaborating with the U.S. Department of Energy (DOE) and DOE's National Renewable Energy Laboratory to evaluate the buses in revenue service. This report summarizes the performance results for the buses through June 2015.

  17. Developing milk industry estimates for dose reconstruction projects

    International Nuclear Information System (INIS)

    Beck, D.M.; Darwin, R.F.

    1991-01-01

    One of the most important contributors to radiation doses from hanford during the 1944-1947 period was radioactive iodine. Consumption of milk from cows that ate vegetation contaminated with iodine is likely the dominant pathway of human exposure. To estimate the doses people could have received from this pathway, it is necessary to reconstruct the amount of milk consumed by people living near Hanford, the source of the milk, and the type of feed that the milk cows ate. This task is challenging because the dairy industry has undergone radical changes since the end of World War 2, and records that document the impact of these changes on the study area are scarce. Similar problems are faced by researchers on most dose reconstruction efforts. The purpose of this work is to document and evaluate the methods used on the Hanford Environmental Dose Reconstruction (HEDR) Project to reconstruct the milk industry and to present preliminary results

  18. Project management for the Virginia power spent fuel storage project

    International Nuclear Information System (INIS)

    Smith, M.

    1992-01-01

    Like Duke Power, Virginia Power has been involved in spent fuel storage expansion studies for a long time - possibly a little longer than Duke Power. Virginia Power's initial studies date back to the late 70s and into the early 80s. Large variety of storage techniques are reviewed including reracking and transshipment. Virginia Power also considered construction a new spent fuel pool. This was one of the options that was considered early on since Virginia Power started this process before any dry storage techniques had been proven. Consolidation of spent fuel is something that was also studied. Finally, construction of dry storage facility was determined to be the technology of choice. They looked a large variety of dry storage technologies and eventually selected dry storage in metal casks at Surry. There are many of reasons why a utility may choose one technology over another. In Virginia Power's situation, additional storage was needed at Surry much earlier than at other utilities. Virginia Power was confronted with selecting a storage technique and having to be a leader in that it was the first U.S. utility to implement a dry storage system

  19. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    BERGMANN, D.W.

    1999-01-01

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  20. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  1. Fuel strategies for natural gas fired cogeneration and IPP projects

    International Nuclear Information System (INIS)

    Gottlieb, J.W.

    1992-01-01

    This paper as published is the outline of a presentation on managing the risk of varying fuel costs as part of a successful fuel strategy for natural gas fired cogeneration and Independent Power Producer (IPP) projects. So long as the fuel cost that electric utilities recover from their ratepayers differs from the fuel costs incurred by IPP and Qualifying Facility (QF) plant operators, the largest variable cost risk of any QF or IPP will continue to be the cost of fuel. Managing that risk is the mission of any successful fuel procurement strategy. Unfortunately, a quick review of the last 20 years in the oil and gas industry reveals dramatic and substantial changes in price and fuel availability that few, if any, industry experts could have predicted in 1971. Recognizing that the fuel cost risk to a QF or IPP investor also spans a 20 year period, the typical term of a QF or IPP power purchase contract, a successful fuel procurement strategy must consider and address the likelihood of future changes. Due to federal and state regulatory changes made from 1978 to 1989, the current structure of the oil and gas industry appears to provide end-users with the tools to improve the manageability of fuel cost risks. QF and IPP developers can choose the type of service they desire and can negotiate most of the contractual elements of that service. Until electric utilities are allowed to flow through their rates the fuel costs incurred by QFs and IPPs, a thorough analysis of the available fuel procurement options prior to development of a QF or IPP will continue to be absolutely necessary

  2. Reactor-specific spent fuel discharge projections, 1984 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Libby, R.A.; Holter, G.M.

    1985-04-01

    The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs

  3. OECD-IAEA Paks Fuel Project. Final Report

    International Nuclear Information System (INIS)

    2010-05-01

    It is important for nuclear power plant designers, operators and regulators to effectively use lessons learned from events occurring at nuclear power plants since, in general, it is impossible to reproduce the event using experimental facilities. In particular, evaluation of the event using accident analysis codes is expected to contribute to improving understanding of phenomena during the events and to facilitate the validation of computer codes through simulation analyses. The information presented in this publication will be of use in future revisions of safety guides on accident analysis. During a fuel crud removal operation on the Paks-2 unit of the Paks nuclear power plant, Hungary on 10 April 2003, several fuel assemblies were severely damaged. The assemblies were being cleaned in a special tank under deep water in a service pit connected to the spent fuel storage pool. The first sign of fuel failures was the detection of some fission gases released from the cleaning tank. Later, visual inspection revealed that most of the 30 fuel assemblies suffered heavy oxidation and fragmentation. The first evaluation of the event showed that the severe fuel damage had been caused by inadequate cooling. The Paks-2 event was discussed in various committees of the OECD Nuclear Energy Agency (OECD/NEA) and of the International Atomic Energy Agency (IAEA). Recommendations were made to undertake actions to improve the understanding of the incident sequence and of the consequence this had on the fuel. It was considered that the Paks-2 event may constitute a useful case for a comparative exercise on safety codes, in particular for models devised to predict fuel damage and potential releases under abnormal cooling conditions and the analyses of the Paks-2 event may provide information which is relevant for in-reactor and spent fuel storage safety evaluations. The OECD-IAEA Paks Fuel Project was established in 2005 as a joint project between the IAEA and the OECD/NEA. The IAEA

  4. Reactor-specific spent fuel discharge projections, 1987-2020

    International Nuclear Information System (INIS)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

  5. Reactor-specific spent fuel discharge projections: 1985 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

  6. Reactor-specific spent fuel discharge projections: 1986 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs

  7. Spent Nuclear Fuel Project Document Management Plan

    International Nuclear Information System (INIS)

    Connor, M.D.; Harizison, G.L.; Rice, W.C.

    1995-12-01

    The SNF Project Document Management Plan identifies and describes the currently available systems and processes for implementing and maintaining an effective document control and records management program. This program governs the methods by which documents are generated, released, distributed, maintained current, retired, and ultimately disposed

  8. The NIOSH Radiation Dose Reconstruction Project: managing technical challenges.

    Science.gov (United States)

    Moeller, Matthew P; Townsend, Ronald D; Dooley, David A

    2008-07-01

    Approximately two years after promulgation of the Energy Employees Occupational Illness Compensation Program Act, the National Institute for Occupational Safety and Health Office of Compensation and Analysis Support selected a contractor team to perform many aspects of the radiation dose reconstruction process. The project scope and schedule necessitated the development of an organization involving a comparatively large number of health physicists. From the initial stages, there were many technical and managerial challenges that required continuous planning, integration, and conflict resolution. This paper identifies those challenges and describes the resolutions and lessons learned. These insights are hopefully useful to managers of similar scientific projects, especially those requiring significant data, technical methods, and calculations. The most complex challenge has been to complete defensible, individualized dose reconstructions that support timely compensation decisions at an acceptable production level. Adherence to applying claimant-favorable and transparent science consistent with the requirements of the Act has been the key to establishing credibility, which is essential to this large and complex project involving tens of thousands of individual stakeholders. The initial challenges included garnering sufficient and capable scientific staff, developing an effective infrastructure, establishing necessary methods and procedures, and integrating activities to ensure consistent, quality products. The continuing challenges include maintaining the project focus on recommending a compensation determination (rather than generating an accurate dose reconstruction), managing the associated very large data and information management challenges, and ensuring quality control and assurance in the presence of an evolving infrastructure. The lessons learned concern project credibility, claimant favorability, project priorities, quality and consistency, and critical

  9. Quality assurance program plan fuel supply shutdown project

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1998-01-01

    This Quality Assurance Program plan (QAPP) describes how the Fuel Supply Shutdown (FSS) project organization implements the quality assurance requirements of HNF-MP-599, Project Hanford Quality Assurance Program Description (QAPD) and the B and W Hanford Company Quality Assurance Program Plan (QAPP), FSP-MP-004. The QAPP applies to facility structures, systems, and components and to activities (e.g., design, procurement, testing, operations, maintenance, etc.) that could affect structures, systems, and components. This QAPP also provides a roadmap of applicable Project Hanford Policies and Procedures (PHPP) which may be utilized by the FSS project organization to implement the requirements of this QAPP

  10. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

    Cherepnin, Yu.S.; Bezzubtsev, V.S.; Gabaraev, B.A.

    2002-01-01

    Positive changes are currently taking place in nuclear power in the world. Power generation at Nuclear Power Plants (NPPs) is increasing and new units construction and completion rates are growing in some of leading countries. Considerable efforts are made for improving the safety of operating NPPs, effective use of nuclear fuel and solving the spent nuclear fuel and radioactive waste problems. Simultaneously, work are undertaken to develop new reactor technologies to reduce the fundamental drawbacks of conventional nuclear power, namely: insufficient safety, spent fuel and waste handling problems, nuclear material proliferation risk and poor economic competitiveness as compared to fossil-fuel energy sources. One the most important events in this field is an international project implemented by three agencies (OECD-IEA, OECD-NEA, IAEA) for comparative evaluation of new projects, development of Generation IV reactors underway in the US in cooperation with a number of Western countries and, finally, the initiative by Russian President V.V. Putin for consolidation the efforts of interested countries under auspices of IAEA to solve the problem of energy support for sustainable development of humankind, radical solution of non-proliferation problems and environmental sanitation of the Planet of Earth. The 44-th General Conference of IAEA in September 2000 supported the Initiative of Russian President and called all interested countries to unite efforts under the Agency's auspices in the International Project on Innovative Nuclear Reactors and Fuel Cycles to consider and select the most acceptable nuclear technologies of the 21-st century with regard for the drawbacks of today's nuclear power. Main objectivities of INPRO: Promotion of the availability of nuclear power for sustainable satisfaction of the energy needs in 21-st century; Consolidation of efforts by all interested INPRO participating countries (both owners and users of technologies) for joint development of

  11. Renewable fuels: Policy effectiveness and project risk

    International Nuclear Information System (INIS)

    Leach, Andrew; Doucet, Joseph; Nickel, Trevor

    2011-01-01

    This paper examines the impact of government policy on the risk profile of a small ethanol production facility. We derive four key results from a simulation model. First, we show that commodity price risk may discourage investment in a project, despite a positive expected rate of return. Second, we show that political uncertainty may have significant impacts on the risk profile of a project. Next, we show that using only production subsidies to attract investors is expensive, since the financial assistance is paid regardless of whether the plant is operating under positive or negative financial conditions. Finally, we show that a capital grant provides a valuable complement to a subsidy, because the grant reduces the amount of value investors must put at risk, and increases their leverage thereby enhancing returns, while the subsidy mitigates commodity price risk. Our results show that compared to a subsidy-only approach, a grant and subsidy combination provides an investment environment with similar downside protection and expected returns for less than 60% of the cost. Further, we show that the two policy tools combined yield a superior investment environment to that created by an equivalent or greater total investment deployed entirely in either of the policy tools without the other. - Research highlights: → We find that government policy may increase both project returns and risk. → We find a policy of capital grants combined with an output price support to be preferred. → Price supports alone will tend to reward those plants which need them the least.

  12. Review on Population Projection Methodology for Radiological Dose Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jang, M. S.; Kang, H. S.; Kim, S. R. [NESS, Daejeon (Korea, Republic of); Hwang, W. T. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yang, Y. H. [KHNP, Daejeon (Korea, Republic of)

    2015-05-15

    Radiation environment report (RER), one of the essential documents in plant operating license or continuous operation license, includes population projection. Population estimates are utilized in determining the collective dose at the operation or restart time of nuclear power plant. Many population projection models are suggested and also under development. We carried out the sensitivity analysis on various population projection models to Daejeon city as a target. Daejeon city showed the increase or decrease in the cross-sectional population, because of the development of Sejong city, Doan new town and etc. We analyzed the population of Daejeon city using statistical ARIMA model and various simple population projection models. It is important to determine the population limit in Modified exponential model but it is not easy. Therefore, the various properties of the area such as the decrease and increase of population, new town development plan, social and natural environment change and etc., should be carefully reviewed to estimate the future population of any area.

  13. Review on Population Projection Methodology for Radiological Dose Assessment

    International Nuclear Information System (INIS)

    Jang, M. S.; Kang, H. S.; Kim, S. R.; Hwang, W. T.; Yang, Y. H.

    2015-01-01

    Radiation environment report (RER), one of the essential documents in plant operating license or continuous operation license, includes population projection. Population estimates are utilized in determining the collective dose at the operation or restart time of nuclear power plant. Many population projection models are suggested and also under development. We carried out the sensitivity analysis on various population projection models to Daejeon city as a target. Daejeon city showed the increase or decrease in the cross-sectional population, because of the development of Sejong city, Doan new town and etc. We analyzed the population of Daejeon city using statistical ARIMA model and various simple population projection models. It is important to determine the population limit in Modified exponential model but it is not easy. Therefore, the various properties of the area such as the decrease and increase of population, new town development plan, social and natural environment change and etc., should be carefully reviewed to estimate the future population of any area

  14. Data base on nuclear power plant dose reduction research projects

    Energy Technology Data Exchange (ETDEWEB)

    Khan, T.A.; Dionne, B.J.; Baum, J.W.

    1985-12-01

    This report contains project information on the research and development activities of the nuclear power industry in the area of dose reduction. It is based on a data base of information set up at the ALARA Center of Brookhaven National Laboratory. One purpose of this report is to draw attention to work in progress and to enable researchers and subscribers to obtain further information from the investigators and project managers. Information is provided on 180 projects, divided according to whether they are oriented to Engineering Research or to Health Physics Technology. The report contains indices on main category, project manager, principal investigator, sponsoring organization, contracting organization, and subject. This is an initial report. It is intended that periodic updates be issued whenever sufficient material has been accumulated.

  15. Data base on nuclear power plant dose reduction research projects

    International Nuclear Information System (INIS)

    Khan, T.A.; Dionne, B.J.; Baum, J.W.

    1985-12-01

    This report contains project information on the research and development activities of the nuclear power industry in the area of dose reduction. It is based on a data base of information set up at the ALARA Center of Brookhaven National Laboratory. One purpose of this report is to draw attention to work in progress and to enable researchers and subscribers to obtain further information from the investigators and project managers. Information is provided on 180 projects, divided according to whether they are oriented to Engineering Research or to Health Physics Technology. The report contains indices on main category, project manager, principal investigator, sponsoring organization, contracting organization, and subject. This is an initial report. It is intended that periodic updates be issued whenever sufficient material has been accumulated

  16. Fuel Cell Balance-of-Plant Reliability Testbed Project

    Energy Technology Data Exchange (ETDEWEB)

    Sproat, Vern [Stark State College of Technology, North Canton, OH (United States); LaHurd, Debbie [Lockheed Martin Corp., Oak Ridge, TN (United States)

    2016-10-29

    Reliability of the fuel cell system balance-of-plant (BoP) components is a critical factor that needs to be addressed prior to fuel cells becoming fully commercialized. Failure or performance degradation of BoP components has been identified as a life-limiting factor in fuel cell systems.1 The goal of this project is to develop a series of test beds that will test system components such as pumps, valves, sensors, fittings, etc., under operating conditions anticipated in real Polymer Electrolyte Membrane (PEM) fuel cell systems. Results will be made generally available to begin removing reliability as a roadblock to the growth of the PEM fuel cell industry. Stark State College students participating in the project, in conjunction with their coursework, have been exposed to technical knowledge and training in the handling and maintenance of hydrogen, fuel cells and system components as well as component failure modes and mechanisms. Three test beds were constructed. Testing was completed on gas flow pumps, tubing, and pressure and temperature sensors and valves.

  17. FY 1991 project plan for the Hanford Environmental Dose Reconstruction Project, Phase 2

    International Nuclear Information System (INIS)

    1991-02-01

    Phase 1 of the Hanford Environmental Dose Reconstruction Project was designed to develop and demonstrate a method for estimating radiation doses people may have received from Hanford Site operations since 1944. The method researchers developed relied on a variety of measured and reconstructed data as input to a modular computer model that generates dose estimates and their uncertainties. As part of Phase 1, researchers used the reconstructed data and computer model to calculate preliminary dose estimates for populations in a limited geographical area and time period. Phase 2, now under way, is designed to evaluate the Phase 1 data and model and improve them to calculate more accurate and precise dose estimates. Phase 2 will also be used to obtain preliminary estimates of two categories of doses: for Native American tribes and for individuals included in the pilot phase of the Hanford Thyroid Disease Study (HTDS). TSP Directive 90-1 required HEDR staff to develop Phase 2 task plans for TSP approval. Draft task plans for Phase 2 were submitted to the TSP at the October 11--12, 1990 public meeting, and, after discussions of each activity and associated budget needs, the TSP directed HEDR staff to proceed with a slate of specific project activities for FY 1991 of Phase 2. This project plan contains detailed information about those activities. Phase 2 is expected to last 15--18 months. In mid-FY 1991, project activities and budget will be reevaluated to determine whether technical needs or priorities have changed. Separate from, but related to, this project plan, will be an integrated plan for the remainder of the project. HEDR staff will work with the TSP to map out a strategy that clearly describes ''end products'' for the project and the work necessary to complete them. This level of planning will provide a framework within which project decisions in Phases 2, 3, and 4 can be made

  18. The Hanford Environmental Dose Reconstruction (HEDR) Project: Technical approach

    International Nuclear Information System (INIS)

    Napier, B.A.; Freshley, M.D.; Gilbert, R.O.; Haerer, H.A.; Morgan, L.G.; Rhoads, R.E.; Woodruff, R.K.

    1990-01-01

    Historical measurements and current assessment techniques are being combined to estimate potential radiation doses to people from radioactive releases to the air, the Columbia River, soils, and ground water at the Hanford Site since 1944. Environmental contamination from these releases has been monitored, at varying levels of detail, for 45 yr. Phase I of the Hanford Environmental Reconstruction Project will estimate the magnitude of potential doses, their areal extends, and their associated uncertainties. The Phase I study area comprises 10 counties in eastern Washington and northern Oregon, within a 100-mi radius of the site, including a stretch of the Columbia River that was most significantly affected. These counties contain a range of projected and measured contaminant levels, environmental exposure pathways, and population groups. Phase I dose estimates are being developed for the periods 1944 through 1947 for air pathways and 1964 through 1966 for river pathways. Important radionuclide/pathway combinations include fission products, such as 131 I, in milk for early atmospheric releases and activation products, such as 32 P and 65 Zn, in fish for releases to the river. Potential doses range over several orders of magnitude within the study area. We will expand the time periods and study are in three successive phases, as warranted by results of Phase I

  19. American Fuel Cell Bus Project Evaluation: Third Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Post, Matthew [National Renewable Energy Lab. (NREL), Golden, CO (United States); Jeffers, Matthew [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-05-01

    This report presents results of the American Fuel Cell Bus (AFCB) Project, a demonstration of fuel cell electric buses operating in the Coachella Valley area of California. The prototype AFCB, which was developed as part of the Federal Transit Administration's (FTA) National Fuel Cell Bus Program, was delivered to SunLine in November 2011 and was put in revenue service in mid-December 2011. Two new AFCBs with an upgraded design were delivered in June/July of 2014 and a third new AFCB was delivered in February 2015. FTA and the AFCB project team are collaborating with the U.S. Department of Energy (DOE) and DOE's National Renewable Energy Laboratory to evaluate the buses in revenue service. This report covers the performance of the AFCBs from July 2015 through December 2016.

  20. Mission Need Statement: Idaho Spent Fuel Facility Project

    Energy Technology Data Exchange (ETDEWEB)

    Barbara Beller

    2007-09-01

    Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

  1. Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1984-09-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy (DOE) radioactive wastes were compiled through December 31, 1983, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 37 years and characteristics of these materials are also presented, consistent with the latest DOE/Energy Information Administration (EIA) or projection of US commercial nuclear power growth and expected defense-related and private industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, airborne waste, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions. 48 figures, 107 tables

  2. Acceptance test procedure for K basins dose reduction project clean and coat equipment

    International Nuclear Information System (INIS)

    Creed, R.F.

    1996-01-01

    This document is the Acceptance Test Procedure (ATP) for the clean and coat equipment designed by Oceaneering Hanford, Inc. under purchase order MDK-XVC-406988 for use in the 105 K East Basin. The ATP provides the guidelines and criteria to test the equipment's ability to clean and coat the concrete perimeter, divider walls, and dummy elevator pit above the existing water level. This equipment was designed and built in support of the Spent Nuclear Fuel, Dose Reduction Project. The ATP will be performed at the 305 test facility in the 300 Area at Hanford. The test results will be documented in WHC-SD-SNF-ATR-020

  3. General criteria for the project of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    1979-01-01

    Recommendations are presented establishing the general criteria for the project of nuclear fuel reprocessing plants to be licensed according to the legislation in effect. They apply to all the plant's systems, components and structures which are important to operation safety and to the public's health and safety. (F.E.) [pt

  4. Expanded spent fuel storage project at Yankee Atomic Electric Plant

    International Nuclear Information System (INIS)

    Chin, S.L.

    1980-01-01

    A detailed discussion on the project at the Yankee Rowe power reactor for expanding the capacity of the at-reactor storage pool by building double-tier storage racks. Various alternatives for providing additional capacity were examined by the operators. Away-from-reactor alternatives included shipment to existing privately owned facilities, a regional independent storage facility, and transshipments to other New England nuclear power plant pools. At-reactor alternatives evaluated included a new pool modification of the existing structure and finally, modification of the spent fuel pit. The establishment of a federal policy precluding transshipment of spent fuel prohibited the use of off-site alternatives. The addition of another pool was too expensive. The possibility of modifying an existing on-site structure required a new safety evaluation by the regulatory group with significant cost and time delays. Therefore, the final alternative - utilizing the existing spent fuel pool with some modification - was chosen due to cost, licensing possibility, no transport requirements, and the fact that the factors involved were mainly under the control of the operator. Modification of the pool was accomplished in phases. In the first phase, a dam was installed in the center of the pool (after the spent fuel was moved to one end). In the second phase, the empty end of the pool was drained and lined with stainless steel and the double-tier rack supports were added. In the third phase, the pool was refilled and the dam was removed. Then the spent fuel was moved into the completed end. In the fourth phase, the dam was replaced and the empty part of the pool was drained. The liner and double-tier rack supports were installed, the pool was refilled, and the dam was removed.The project demonstrated that the modification of existing spent fuel fuel pools for handling double-tier fuel racks is a viable solution for increasing the storage capacity at the reactor

  5. Data base on nuclear power plant dose reduction research projects

    Energy Technology Data Exchange (ETDEWEB)

    Khan, T.A.; Baum, J.W.

    1986-10-01

    Staff at the ALARA Center of Brookhaven National Laboratory have established a data base of information about current research that is likely to result in lower radiation doses to workers. The data base, concerned primarily with nuclear power generation, is part of a project that the ALARA Center is carrying out for the Nuclear Regulatory Commission. This report describes its current status. A substantial amount of research on reducing occupational exposure is being done in the US and abroad. This research is beginning to have an impact on the collective dose expenditures at nuclear power plants. The collective radiation doses in Europe, Japan, and North America all show downward trends. A large part of the research in the US is either sponsored by the nuclear industry through joint industry organizations such as EPRI and ESEERCO or is done by individual corporations. There is also significant participation by smaller companies. The main emphasis of the research on dose reduction is on engineering approaches aimed at reducing radiation fields or keeping people out of high-exposure areas by using robotics. Effective ALARA programs are also underway at a large number of nuclear plants. Additional attention should be given to non-engineering approaches to dose reduction, which are potentially very useful and cost effective but require quantitative study and analysis based on data from nuclear power plants. 9 refs., 1 fig.

  6. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    International Nuclear Information System (INIS)

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-01

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of ∼10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate

  7. Guidance on internal dose assessments from monitoring data (Project IDEAS)

    International Nuclear Information System (INIS)

    Doerfel, H.; Andrasi, A.; Bailey, M.; Berkovski, V.; Castellani, M.; Hurtgen, C.; Jourdain, R.; Le Guen, B.

    2003-01-01

    Several international intercomparison exercises on intake and internal dose assessments from monitoring data led to the conclusion that the results calculated by different participants varied significantly mainly to the broad variety of methods and assumptions applied in the assessment procedure. Based on these experiences the need of harmonisation of the procedures has been formulated as an EU research project under the 5th Framework Programme, with the aim of developing general guidelines for standardising assessments of intakes and internal doses. In the IDEAS project, eight institutions from seven European countries are participating, also using inputs from internal dosimetry professionals from across Europe to ensure broad consensus in the outcome of the project. To ensure that the guidelines are applicable to a wide range of practical situations, the first step will be to compile a database on well documented cases of internal contamination. In parallel, an improved version of existing software will be developed and distributed to the partners for further use. Many cases from the database will be evaluated independently by more partners using the same software and the results will be discussed and the draft guidelines prepared. The guidelines will then be revised and refined on the basis of the experiences and discussions of two workshops, and an inter-comparison exercise organised in the frame of the project which will be open to all internal dosimetry professionals. (author)

  8. Hanford Environmental Dose Reconstruction Project independent direction and oversight

    International Nuclear Information System (INIS)

    Blazek, M.L.; Power, M.

    1991-01-01

    Hanford was selected in 1942 as one of the sites for the Manhattan Project. It produced plutonium for one of the world's first nuclear weapons. The US Department of Energy (DOE) and its predecessors continued to make plutonium for nuclear weapons at Hanford for more than four decades. In the early days of Hanford operations, radioactive materials routinely were released to the environment by many processes. The DOE disclosed documents about these releases in 1986. In 1987, Washington, Oregon, and regional Indian tribes gathered an independent panel of experts. This group recommended dose reconstruction and health effects feasibility studies. Later that year, DOE hired Battelle Pacific Northwest Laboratory (PNL) to reconstruct potential public radiation doses from Hanford's past releases of radioactive material. The DOE agreed with the states and tribes that project direction would come from an independent technical steering panel (TSP). This approach was critical to gain public credibility for the project and the science. The TSP directs the project and makes policy. That is now clear - but, it was hard-earned. Conducting science in an open public process is new, challenging, and clearly worthwhile. The panel's product is good science that is believed and accepted by the public - our client

  9. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

  10. DIMETHYL ETHER (DME)-FUELED SHUTTLE BUS DEMONSTRATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Elana M. Chapman; Shirish Bhide; Jennifer Stefanik; Howard Glunt; Andre L. Boehman; Allen Homan; David Klinikowski

    2003-04-01

    The objectives of this research and demonstration program are to convert a campus shuttle bus to operation on dimethyl ether, a potential ultra-clean alternative diesel fuel. To accomplish this objective, this project includes laboratory evaluation of a fuel conversion strategy, as well as, field demonstration of the DME-fueled shuttle bus. Since DME is a fuel with no lubricity (i.e., it does not possess the lubricating quality of diesel fuel), conventional fuel delivery and fuel injection systems are not compatible with dimethylether. Therefore, to operate a diesel engine on DME one must develop a fuel-tolerant injection system, or find a way to provide the necessary lubricity to the DME. In this project, they have chosen the latter strategy in order to achieve the objective with minimal need to modify the engine. The strategy is to blend DME with diesel fuel, to obtain the necessary lubricity to protect the fuel injection system and to achieve low emissions. The bulk of the efforts over the past year were focused on the conversion of the campus shuttle bus. This process, started in August 2001, took until April 2002 to complete. The process culminated in an event to celebrate the launching of the shuttle bus on DME-diesel operation on April 19, 2002. The design of the system on the shuttle bus was patterned after the system developed in the engine laboratory, but also was subjected to a rigorous failure modes effects analysis with help from Dr. James Hansel of Air Products. The result of this FMEA was the addition of layers of redundancy and over-pressure protection to the system on the shuttle bus. The system became operation in February 2002. Preliminary emissions tests and basic operation of the shuttle bus took place at the Pennsylvania Transportation institute's test track facility near the University Park airport. After modification and optimization of the system on the bus, operation on the campus shuttle route began in early June 2002. However, the

  11. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared to a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output

  12. Dose rate analyses for fast reactor fuel manufacturers

    International Nuclear Information System (INIS)

    Smith, R.C.; Strode, J.N.; Brackenbush, L.W.; Faust, L.G.

    1976-01-01

    An early appraisal of the radiation exposure situation in the fabrication of plutonium enriched mixed-oxide fuels for fast reactors is presented. Radiation data are presented on fuel process operations measured under actual operating conditions using plutonium containing up to 19 wt percent 240 Pu

  13. Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1983-09-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled through December 31, 1982, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 40 years and characteristics of these materials are also presented, based on the latest DOE/EIA projection of US commercial nuclear power growth and expected defense-related and industrial and institutional activities. Materials considered, on a chapter-by-chapter bases, are: spent fuel, high-level waste, transuranic waste, low-level waste, active uranium mill tailings, airborne waste, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions. One chapter gives broad, summary data on the costs of spent fuel and radioactive waste management and disposal to provide an economic perspective. This chapter is not intended as a definitive guide, but it is a source of reasonable, order-of-magnitude costs and also provides references to more-detailed and scenario-specific studies. An appendix on generic flowsheets and source terms used for the projections is also included

  14. Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

    International Nuclear Information System (INIS)

    Cooper, T.D.; Pajunen, A.L.

    1998-01-01

    The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site's K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients

  15. Doses of the staff during the spent fuel assemblies transportation and storage in Nuhmos 56V concrete system

    International Nuclear Information System (INIS)

    Atoyan, V.; Muradyan, A.

    2003-01-01

    The NUHMOS 56V concrete system provides long-term interim storage (50 years) for spent fuel assemblies, which have been out of the reactor for a sufficient period of time. It consists from horizontal storage modules. The fuel assemblies are confined in a helium atmosphere by a canister containment pressure vessel. The canister is protected and shielded by a massive reinforced concrete module. Decay heat is removed from the canister and concrete module by a passive natural draft convection ventilation system. The project of storage does not foresee the radiation monitoring inside of building and around it. But we provided and realize the radiation monitoring program around storage, it includes tree phases: - determination the zero background around the building before storage put in exploiting; - monitoring of the radioactive particles in air (additional aspiration plant); dose rate monitoring by portable dosimeters and soil monitoring during the process of the fuel storage; - constantly after the completion the fuel storage process - monitoring of the radioactive particles in air (additional aspiration plant); dose rate monitoring by portable dosimeters, and soil monitoring. Also designed the dose rate monitoring by the dosimeter RME 3 with the transfer of data by radio channel to central monitor. The canistered spent fuel assemblies are transferred from the plant's spent fuel pool to the concrete storage modules in a transfer cask. The cask is aligned with the storage module and the canister and inserted into the module by means of a hydraulic ram. The system is a totally passive installation that is designed to provide shielding and safe confinement of spent fuel for a range of postulated accident conditions and natural phenomena. (authors)

  16. Calculation of nuclide inventory, decay power, activity and dose rates for spent nuclear fuel

    International Nuclear Information System (INIS)

    Haakansson, Rune

    2000-03-01

    The nuclide inventory was calculated for a BWR and a PWR fuel element, with burnups of 38 and 55 MWd/kg uranium for the BWR fuel, and 42 and 60 MWd/kg uranium for the PWR fuel. The calculations were performed for decay times of up to 300,000 years. Gamma and neutron dose rates have been calculated at a distance of 1 m from a bare fuel element and outside the spent fuel canister. The calculations were performed using the CASMO-4 code

  17. Hanford K basins spent nuclear fuel project update

    International Nuclear Information System (INIS)

    Williams, N.H.; Hudson, F.G.

    1997-07-01

    Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building

  18. Sweet Sorghum Alternative Fuel and Feed Pilot Project

    Energy Technology Data Exchange (ETDEWEB)

    Slack, Donald C. [Univ. of Arizona, Tucson, AZ (United States). Agricultural and Biosystems Engineering Dept.; Kaltenbach, C. Colin [Univ. of Arizona, Tucson, AZ (United States)

    2013-07-30

    The University of Arizona undertook a “pilot” project to grow sweet sorghum on a field scale (rather than a plot scale), produce juice from the sweet sorghum, deliver the juice to a bio-refinery and process it to fuel-grade ethanol. We also evaluated the bagasse for suitability as a livestock feed and as a fuel. In addition to these objectives we evaluated methods of juice preservation, ligno-cellulosic conversion of the bagasse to fermentable sugars and alternative methods of juice extraction.

  19. A Renewably Powered Hydrogen Generation and Fueling Station Community Project

    Science.gov (United States)

    Lyons, Valerie J.; Sekura, Linda S.; Prokopius, Paul; Theirl, Susan

    2009-01-01

    The proposed project goal is to encourage the use of renewable energy and clean fuel technologies for transportation and other applications while generating economic development. This can be done by creating an incubator for collaborators, and creating a manufacturing hub for the energy economy of the future by training both white- and blue-collar workers for the new energy economy. Hydrogen electrolyzer fueling stations could be mass-produced, shipped and installed in collaboration with renewable energy power stations, or installed connected to the grid with renewable power added later.

  20. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  1. Oak Ridge Dose Reconstruction Project Summary Report; Reports of the Oak Ridge Dose Reconstruction, Vol. 7

    Energy Technology Data Exchange (ETDEWEB)

    Thomas E. Widner; et. al.

    1999-07-01

    In the early 1990s, concern about the Oak Ridge Reservation's past releases of contaminants to the environment prompted Tennessee's public health officials to pursue an in-depth study of potential off-site health effects at Oak Ridge. This study, the Oak Ridge dose reconstruction, was supported by an agreement between the U.S. Department of Energy (DOE) and the State of Tennessee, and was overseen by a 12-member panel of individuals appointed by Tennessee's Commissioner of Health. The panel requested that the principal investigator for the project prepare the following report, ''Oak Ridge Dose Reconstruction Project Summary Report,'' to serve the following purposes: (1) summarize in a single, less technical report, the methods and results of the various investigations that comprised the Phase II of the dose reconstruction; (2) describe the systematic searching of classified and unclassified historical records that was a vital component of the project; and (3) summarize the less detailed, screening-level assessments that were performed to evaluate the potential health significance of a number of materials, such a uranium, whose priority did not require a complete dose reconstruction effort. This report describes each major step of the dose reconstruction study: (1) the review of thousands of historical records to obtain information relating to past operations at each facility; (2) estimation of the quantity and timing of releases of radioiodines from X-10, of mercury from Y-12, of PCB's from all facilities, and of cesium-137 and other radionuclides from White Oak Creek; (3) evaluation of the routes taken by these contaminants through the environment to nearby populations; and (4) estimation of doses and health risks to exposed groups. Calculations found the highest excess cancer risks for a female born in 1952 who drank goat milk; the highest non-cancer health risk was for children in a farm family exposed to PCBs in and near

  2. Oak Ridge Dose Reconstruction Project Summary Report; Reports of the Oak Ridge Dose Reconstruction, Vol. 7

    International Nuclear Information System (INIS)

    Widner, Thomas E.; email = twidner@jajoneses.com

    1999-01-01

    In the early 1990s, concern about the Oak Ridge Reservation's past releases of contaminants to the environment prompted Tennessee's public health officials to pursue an in-depth study of potential off-site health effects at Oak Ridge. This study, the Oak Ridge dose reconstruction, was supported by an agreement between the U.S. Department of Energy (DOE) and the State of Tennessee, and was overseen by a 12-member panel of individuals appointed by Tennessee's Commissioner of Health. The panel requested that the principal investigator for the project prepare the following report, ''Oak Ridge Dose Reconstruction Project Summary Report,'' to serve the following purposes: (1) summarize in a single, less technical report, the methods and results of the various investigations that comprised the Phase II of the dose reconstruction; (2) describe the systematic searching of classified and unclassified historical records that was a vital component of the project; and (3) summarize the less detailed, screening-level assessments that were performed to evaluate the potential health significance of a number of materials, such a uranium, whose priority did not require a complete dose reconstruction effort. This report describes each major step of the dose reconstruction study: (1) the review of thousands of historical records to obtain information relating to past operations at each facility; (2) estimation of the quantity and timing of releases of radioiodines from X-10, of mercury from Y-12, of PCB's from all facilities, and of cesium-137 and other radionuclides from White Oak Creek; (3) evaluation of the routes taken by these contaminants through the environment to nearby populations; and (4) estimation of doses and health risks to exposed groups. Calculations found the highest excess cancer risks for a female born in 1952 who drank goat milk; the highest non-cancer health risk was for children in a farm family exposed to PCBs in and near East Fork Poplar Creek. More detailed

  3. Transport fuel demand responses to fuel price and income projections : Comparison of integrated assessment models

    NARCIS (Netherlands)

    Edelenbosch, O. Y.; van Vuuren, Detlef; Bertram, C.; Carrara, S.; Emmerling, J.; Daly, H.; Kitous, A.; McCollum, D. L.; Saadi Failali, N.

    Income and fuel price pathways are key determinants in projections of the energy system in integrated assessment models. In recent years, more details have been added to the transport sector representation in these models. To better understand the model dynamics, this manuscript analyses transport

  4. Some aspects concerning the implementation of a fuel technology project

    International Nuclear Information System (INIS)

    Andreescu, N.; Alecu, M.; Mirion, I.

    1977-01-01

    The nuclear power programme in Romania envisages that until 1990 there will be installed about 6000 MWe in nuclear power plants. In order to put into practice such a nuclear programme there will be necessary high investments, possible to be achieved only by the ever increasing participation of the Romanian industry. With a view to this purpose, the Romanian authorities pay great attention to the research and development of the nuclear fuel manufacturing technology. Some research started in 1968-1969 and was intensified later in 1971 when the Institute for Nuclear Technology was founded and in 1972 when the IAEA-UNDP programme ''Development of Nuclear Technology in Romania'' started. This programme was conceived to deal with; 1. technology of UO2 powder and pellet fabrication; 2. manufacturing technology of fuel rods and bundle; 3. irradiation test of fuel rods; 4. development of various activities connected to fuel technology (thermal transfer loops, corrosion tests, neutronic, thermal and hydrodynamical calculations). Within the IAEA-UNDP project a demonstration facility was installed at INT where a great number of the works, resulting from the above mentioned directions, were performed. As a result of these works, at the end of 1975 in the demonstration facility there were manufactured in a reproducible way fuel rods according to the required specifications. The paper further presents the adopted irradiation testing programme, the out-of-pile testing programme, as well as some performances obtained during the different phases of the whole project. There have been conceived and manufactured some equipment meant for fabrication, tests, or for current control. The paper also shows some aspects connected to the personnel formation, as well as some aspects that will have to be solved in order to make possible the step from the demonstration facility to a fuel plant

  5. Texas LPG fuel cell development and demonstration project

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2004-07-26

    The State Energy Conservation Office has executed its first Fuel Cell Project which was awarded under a Department of Energy competitive grant process. The Texas LPG Fuel Processor Development and Fuel Cell Demonstration Program is a broad-based public/private partnership led by the Texas State Energy Conservation Office (SECO). Partners include the Alternative Fuels Research and Education Division (AFRED) of the Railroad Commission of Texas; Plug Power, Inc., Latham, NY, UOP/HyRadix, Des Plaines, IL; Southwest Research Institute (SwRI), San Antonio, TX; the Texas Natural Resource Conservation Commission (TNRCC), and the Texas Department of Transportation (TxDOT). The team proposes to mount a development and demonstration program to field-test and evaluate markets for HyRadix's LPG fuel processor system integrated into Plug Power's residential-scale GenSys(TM) 5C (5 kW) PEM fuel cell system in a variety of building types and conditions of service. The program's primary goal is to develop, test, and install a prototype propane-fueled residential fuel cell power system supplied by Plug Power and HyRadix in Texas. The propane industry is currently funding development of an optimized propane fuel processor by project partner UOP/HyRadix through its national checkoff program, the Propane Education and Research Council (PERC). Following integration and independent verification of performance by Southwest Research Institute, Plug Power and HyRadix will produce a production-ready prototype unit for use in a field demonstration. The demonstration unit produced during this task will be delivered and installed at the Texas Department of Transportation's TransGuide headquarters in San Antonio, Texas. Simultaneously, the team will undertake a market study aimed at identifying and quantifying early-entry customers, technical and regulatory requirements, and other challenges and opportunities that need to be addressed in planning commercialization of the units

  6. Memorandum concerning the European project of dose passport (dosimetry booklet)

    International Nuclear Information System (INIS)

    2013-01-01

    In fact the European project represents the implementation in European law of the 90/641 EURATOM directive that proposed a common European system for the follow-up of the occupational irradiation of workers. The EURATOM directive recommends a computer system while the European project proposes to write down information in a simple booklet. Some experts highlight the fact that it would be easier and more reliable to upgrade a computer file than a booklet and that the information must be available in different European languages. The experts recommend that the European countries must agree on what information would be compulsory, and on an accurate definition of the radiation dose we have to report and on how to measure it in order to have a consistent system throughout Europe. (A.C.)

  7. Revue of some dosimetry and dose assessment European projects

    International Nuclear Information System (INIS)

    Bolognese-Milsztajn, T.; Frank, D.; Lacoste, V.; Pihet, P.

    2006-01-01

    Full text of publication follows: Within the 5. Framework Programme of the European Commission several project dealing with dosimetry and dose assessment for internal and external exposure have been supported. A revue of the results of some of them is presented in this paper. The EURADOS network which involved 50 dosimetry institutes in EUROPE has coordinated the project DOSIMETRY NETWORK: the main results achieved within this action are the following: - The release on the World Wide Web of the EURADOS Database of Dosimetry Research Facilities; - The realisation of the report 'Harmonization of Individual Monitoring (IM) in Europe'; - The continuation of the intercomparisons programme of environmental radiation monitoring systems; - The realisation of the report Cosmic radiation exposure of aircraft crew. The EVIDOS project aimed at evaluating state of the art dosimetry techniques in representative workplaces of the nuclear industry with complex mixed neutron-photon radiation fields. This paper summarises the main findings from a practical point of view. Conclusions and recommendations will be given concerning characterisation of radiation fields, methods to derive radiation protection quantities and dosimeters results. The IDEA project aimed to improve the assessment of incorporated radionuclides through developments of advanced in-vivo and bioassay monitoring techniques and making use of such enhancements for improvements in routine monitoring. The primary goal was to categorize those new developments regarding their potential and eligibility for the routine monitoring community. The costs of monitoring for internal exposures in the workplace are usually significantly greater than the equivalent costs for external exposures. There is therefore a need to ensure that resources are employed with maximum effectiveness. The EC-funded OMINEX (Optimisation of Monitoring for Internal Exposure) project has developed methods for optimising the design and implementation of

  8. Spent fuel and radioactive-waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1982-10-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 40 years, and characteristics of these materials are also presented, based on a present DOE/EIA projection of US commercial nuclear power growth and expected defense-related and industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, remedial action waste, active uranium mill tailings, airborne waste, and decommissioning. For each category, current and projected inventories are given through the year 2020. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for the present inventories of inactive uranium mill tailings. For each waste category the radioactivity and thermal power are calculated. Isotopic compositions and cost data are given for each waste type and for spent fuel

  9. Projections of Full-Fuel-Cycle Energy and Emissions Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Coughlin, Katie [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2013-01-01

    To accurately represent how conservation and efficiency policies affect energy demand, both direct and indirect impacts need to be included in the accounting. The indirect impacts are defined here as the resource savings that accrue over the fuel production chain, which when added to the energy consumed at the point of use, constitute the full-fuel- cycle (FFC) energy. This paper uses the accounting framework developed in (Coughlin 2012) to calculate FFC energy metrics as time series for the period 2010-2040. The approach is extended to define FFC metrics for the emissions of greenhouse gases (GHGs) and other air-borne pollutants. The primary focus is the types of energy used in buildings and industrial processes, mainly natural gas and electricity. The analysis includes a discussion of the fuel production chain for coal, which is used extensively for electric power generation, and for diesel and fuel oil, which are used in mining, oil and gas operations, and fuel distribution. Estimates of the energy intensity parameters make use of data and projections from the Energy Information Agency’s National Energy Modeling System, with calculations based on information from the Annual Energy Outlook 2012.

  10. 78 FR 13315 - Bridger-Teton National Forest; Wyoming; Teton to Snake Fuels Management Project

    Science.gov (United States)

    2013-02-27

    ... Fuels Management Project AGENCY: Forest Service, USDA. ACTION: Notice of intent to prepare an...) to document the potential effects of the Teton to Snake Fuels Management Project. The analysis will... Caribou-Targhee National Forest. The Teton to Snake Fuels Management Project was previously scoped and...

  11. Study on transport safety of refresh MOX fuel. Radiation dose from package hypothetically submerged into sea

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Suzuki; Hiroshi; Saegusa, Toshiari; Maruyama, Koki; Ito, Chihiro; Watabe, Naoto

    1999-01-01

    The sea transport of fresh MOX fuel from Europe to Japan is under planning. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel will be transported safely on the sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unexpected reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

  12. Spent fuel transportation on highways: the radioactive dose to the traffic

    International Nuclear Information System (INIS)

    Yadigaroglu, G.

    1975-01-01

    The radioactive exposure of the traffic moving on the same highway as spent fuel shipments has been neglected in the past. Methods developed for calculating peak exposures, the number of individuals receiving a dose in excess of a certain limiting value, and the cumulative population doses for the occupants of the vehicles under a variety of highway and accident conditions allow comparisons to the corresponding stationary-population doses. Consideration of both routine direct-radiation exposures and accidental releases indicates that the traffic doses can be of equal or greater importance than the stationary-population doses

  13. Fuel cell propulsion for urban duty vehicles: Bavarian fuel cell bus project

    International Nuclear Information System (INIS)

    Wurster, R.; Altmann, M.; Sillat, D.; Kalk, K. W.; Hammerschmidt, A.; Stuehler, W.; Holl, E.

    1998-01-01

    Following a feasibility study and a detailed specification phase, the realization of a fuel cell city bus prototype was started in autumn 1996. The project is a joint development effort of Siemens, MAN and Linde, which receives a 50 % funding by the Bavarian State Ministry for Economic Affairs, Transport and Technology (BStMWVT) in the context of the Hydrogen Initiative Bavaria. An MAN low-floor bus will be equipped with the components for a fuel cell drive system. The PEM fuel cell is developed by the power generation division of Siemens. Four fuel cell modules deliver a total electrical output of 120 kW to the two electric motors, which are linked by a summation gearbox by the Siemens Transportation Systems Division. MAN Technologie AG is responsible for the compressed hydrogen storage system allowing for a driving range of more than 250 km, while Linde AG takes care of the hydrogen periphery and delivers the hydrogen for the test operation scheduled for the beginning of the year 2000. Project coordination is done by Ludwig-Boelkow System-technik GmbH. The project is divided into four phases. The conceptual design phase is scheduled to last until the end of 1997. The partly overlapping system integration phase will end in the first quarter of 1999. The subsequent test and commissioning phase will prepare the test operation at the beginning of 2000 with a bus operator yet to be defined. (author)

  14. High dose stainless steel swelling data on interior and peripheral oxide fuel pins

    International Nuclear Information System (INIS)

    Boltax, A.; Foster, J.P.; Nayak, U.P.

    1983-01-01

    High dose (2 x 10 23 n/cm 2 , E > 0.1 Mev) swelling data obtained on 20% cold-worked AISI 316 stainless steel (N-lot) cladding from mixed-oxide fuel pins show large differences in swelling incubation dose due to pre-incubation dose temperature changes. Circumferential swelling variations of 1.5 to 4 times were found in peripheral fuel pin cladding which experienced 30 to 60 deg C temperature changes due to movement in a temperature gradient. Consideration is given to the implications of these results to low swelling materials development and core design. (author)

  15. Temperature and neutron dose rate measurements at a spent fuel shipping cask

    International Nuclear Information System (INIS)

    Krause, F.

    1982-01-01

    Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

  16. Messiah College Biodiesel Fuel Generation Project Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Zummo, Michael M; Munson, J; Derr, A; Zemple, T; Bray, S; Studer, B; Miller, J; Beckler, J; Hahn, A; Martinez, P; Herndon, B; Lee, T; Newswanger, T; Wassall, M

    2012-03-30

    Many obvious and significant concerns arise when considering the concept of small-scale biodiesel production. Does the fuel produced meet the stringent requirements set by the commercial biodiesel industry? Is the process safe? How are small-scale producers collecting and transporting waste vegetable oil? How is waste from the biodiesel production process handled by small-scale producers? These concerns and many others were the focus of the research preformed in the Messiah College Biodiesel Fuel Generation project over the last three years. This project was a unique research program in which undergraduate engineering students at Messiah College set out to research the feasibility of small-biodiesel production for application on a campus of approximately 3000 students. This Department of Energy (DOE) funded research program developed out of almost a decade of small-scale biodiesel research and development work performed by students at Messiah College. Over the course of the last three years the research team focused on four key areas related to small-scale biodiesel production: Quality Testing and Assurance, Process and Processor Research, Process and Processor Development, and Community Education. The objectives for the Messiah College Biodiesel Fuel Generation Project included the following: 1. Preparing a laboratory facility for the development and optimization of processors and processes, ASTM quality assurance, and performance testing of biodiesel fuels. 2. Developing scalable processor and process designs suitable for ASTM certifiable small-scale biodiesel production, with the goals of cost reduction and increased quality. 3. Conduct research into biodiesel process improvement and cost optimization using various biodiesel feedstocks and production ingredients.

  17. Hanford K Basins spent nuclear fuels project update

    International Nuclear Information System (INIS)

    Hudson, F.G.

    1997-01-01

    Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed

  18. Hanford K Basins spent nuclear fuels project update

    Energy Technology Data Exchange (ETDEWEB)

    Hudson, F.G.

    1997-10-17

    Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed.

  19. Krakow clean fossil fuels and energy efficiency project

    Energy Technology Data Exchange (ETDEWEB)

    Butcher, T.A.; Pierce, B.L. [Brookhaven National Lab., Upton, NY (United States)

    1995-11-01

    The Support for Eastern European Democracy (SEED) Act of 1989 directed the U.S. Department of Energy (DOE) to undertake an equipment assessment project aimed at developing the capability within Poland to manufacture or modify industrial-scale combustion equipment to utilize fossil fuels cleanly. This project is being implemented in the city of Krakow as the `Krakow Clean Fossil Fuels and Energy Efficiency Project.` Funding is provided through the U.S. Agency for International Development (AID). The project is being conducted in a manner that can be generalized to all of Poland and to the rest of Eastern Europe. The historic city of Krakow has a population of 750,000. Almost half of the heating energy used in Krakow is supplied by low-efficiency boilerhouses and home coal stoves. Within the town, there are more than 1,300 local boilerhouses and 100,000 home stoves. These are collectively referred to as the `low emission sources` and they are the primary sources of particulates and hydrocarbon emissions in the city and major contributors of sulfur dioxide and carbon monoxide.

  20. Dose and dose commitment calculations from groundwaterborne radio-active elements released from a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Bergstroem, U.

    1983-05-01

    The turnover of radioactive matter entering the biosphere with groundwater has been studied with regard to exposure and doses to critical groups and populations. Two main recipients, a well and a lake, have been considered for the inflow of groundwaterborne nuclides. Mathematical models of a set of coupled ecosystems on regional, intermediate and global levels have been used for calculations of doses. The intermediate system refers to the Baltic Sea. The mathematical treatment of the model is based upon compartment theory with first order kinetics and also includes products in decay chains. The time-dependent exposures have been studied for certain long-lived nuclides of radiological interest in waste from disposed fuel. Dose and dose commitment have been calculated for different episodes for inflow to the biosphere. (author)

  1. Current status and prospects on Rokkasho nuclear fuel cycle project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuo

    2003-01-01

    JNFL has been established aiming at fulfillment of Nuclear Fuel Cycle, as well as to contribute to the long-term and stable supply of nuclear power in Japan. 'Uranium Enrichment Plant' with its production of 1,050 SWU/y and planned to be expand to 1,500 SWU/y, 'Low Level Radioactive Waste Disposal Center' with 150,000/200 l drums stored, out of its 400,000 drums capacity, and 'Vitrified Waste Storage Center' with 760 canisters stored, out of its 1440 canisters capacity, are already in its operation. It is now preparing for the operation of '800 t/y Reprocessing Plant' and construction of '130t HM/y MOX Fuel Fabrication Plant'. As for the Reprocessing Plant, 780t of spent fuels has been already received and stored in the storage pools. Main plant is now in the course of test operation and planned to start the commercial operation by July 2006. Due to some defects found during the course of its construction, JNFL is now reviewing the Total Quality Assurance Structure to improve and reinforce its system. And for the MOX Fuel Fabrication Plant, activities towards obtaining the local autonomy's agreement for the construction are being made energetically. It is essential to obtain the good understanding of the public community to promote these projects successfully; JNFL is putting its best efforts to dispatch all the necessary information to the public in a timely manner. (author)

  2. DIMETHYL ETHER (DME)-FUELED SHUTTLE BUS DEMONSTRATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Elana M. Chapman; Shirish Bhide; Jennifer Stefanik; Howard Glunt; Andre L. Boehman; Allen Homan; David Klinikowski

    2003-04-01

    The objectives of this research and demonstration program are to convert a campus shuttle bus to operation on dimethyl ether, a potential ultra-clean alternative diesel fuel. To accomplish this objective, this project includes laboratory evaluation of a fuel conversion strategy, as well as, field demonstration of the DME-fueled shuttle bus. Since DME is a fuel with no lubricity (i.e., it does not possess the lubricating quality of diesel fuel), conventional fuel delivery and fuel injection systems are not compatible with dimethyl ether. Therefore, to operate a diesel engine on DME one must develop a fuel-tolerant injection system, or find a way to provide the necessary lubricity to the DME. In this project, they have chosen the latter strategy in order to achieve the objective with minimal need to modify the engine. Their strategy is to blend DME with diesel fuel, to obtain the necessary lubricity to protect the fuel injection system and to achieve low emissions. The bulk of the efforts over the past year were focused on the conversion of the campus shuttle bus. This process, started in August 2001, took until April 2002 to complete. The process culminated in an event to celebrate the launching of the shuttle bus on DME-diesel operation on April 19, 2002. The design of the system on the shuttle bus was patterned after the system developed in the engine laboratory, but also was subjected to a rigorous failure modes effects analysis (FMEA, referred to by Air Products as a ''HAZOP'' analysis) with help from Dr. James Hansel of Air Products. The result of this FMEA was the addition of layers of redundancy and over-pressure protection to the system on the shuttle bus. The system became operational in February 2002. Preliminary emissions tests and basic operation of the shuttle bus took place at the Pennsylvania Transportation Institute's test track facility near the University Park airport. After modification and optimization of the system on

  3. Analysis of gamma irradiator dose rate using spent fuel elements with parallel configuration

    International Nuclear Information System (INIS)

    Setiyanto; Pudjijanto MS; Ardani

    2006-01-01

    To enhance the utilization of the RSG-GAS reactor spent fuel, the gamma irradiator using spent fuel elements as a gamma source is a suitable choice. This irradiator can be used for food sterilization and preservation. The first step before realization, it is necessary to determine the gamma dose rate theoretically. The assessment was realized for parallel configuration fuel elements with the irradiation space can be placed between fuel element series. This analysis of parallel model was choice to compare with the circle model and as long as possible to get more space for irradiation and to do manipulation of irradiation target. Dose rate calculation were done with MCNP, while the estimation of gamma activities of fuel element was realized by OREGEN code with 1 year of average delay time. The calculation result show that the gamma dose rate of parallel model decreased up to 50% relatively compared with the circle model, but the value still enough for sterilization and preservation. Especially for food preservation, this parallel model give more flexible, while the gamma dose rate can be adjusted to the irradiation needed. The conclusion of this assessment showed that the utilization of reactor spent fuels for gamma irradiator with parallel model give more advantage the circle model. (author)

  4. Fuel integrity project: analysis of light water reactor fuel rods test results

    Energy Technology Data Exchange (ETDEWEB)

    Dallongeville, M.; Werle, J. [COGEMA Logistics (AREVA Group) (France); McCreesh, G. [BNFL Nuclear Sciences and Technology Services (United Kingdom)

    2004-07-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  5. Fuel integrity project: analysis of light water reactor fuel rods test results

    International Nuclear Information System (INIS)

    Dallongeville, M.; Werle, J.; McCreesh, G.

    2004-01-01

    BNFL Nuclear Sciences and Technology Services and COGEMA LOGISTICS started in the year 2000 a joint project known as FIP (Fuel Integrity Project) with the aim of developing realistic methods by which the response of LWR fuel under impact accident conditions could be evaluated. To this end BNFL organised tests on both unirradiated and irradiated fuel pin samples and COGEMA LOGISTICS took responsibility for evaluating the test results. Interpretation of test results included simple mechanical analysis as well as simulation by Finite Element Analysis. The first tests that were available for analysis were an irradiated 3 point bending commissioning trial and a lateral irradiated hull compression test, both simulating the loading during a 9 m lateral regulatory drop. The bending test span corresponded roughly to a fuel pin intergrid distance. The outcome of the test was a failure starting at about 35 mm lateral deflection and a few percent of total deformation. Calculations were carried out using the ANSYS code employing a shell and brick model. The hull lateral compaction test corresponds to a conservative compression by neighbouring pins at the upper end of the fuel pin. In this pin region there are no pellets inside. The cladding broke initially into two and later into four parts, all of which were rather similar. Initial calculations were carried out with LS-DYNA3D models. The models used were optimised in meshing, boundary conditions and material properties. The calculation results compared rather well with the test data, in particular for the detailed ANSYS approach of the 3 point bending test, and allowed good estimations of stresses and deformations under mechanical loading as well as the derivation of material rupture criteria. All this contributed to the development of realistic numerical analysis methods for the evaluation of LWR fuel rod behaviour under both normal and accident transport conditions. This paper describes the results of the 3 point bending

  6. Leading fuel cell projects ``PEM``; Brennstoffzellen-Leitprojekte ``PEM``

    Energy Technology Data Exchange (ETDEWEB)

    Isenberg, G

    1996-12-31

    Polymer electrolyte diaphragm fuel cells (PEMFC) are distinguished by high specific power densities which are necessary in the field of electrical drives, high efficiencies and no or very low emission of harmful substances, depending on the fuel. In the context of a leading project supported by BMBF, the development of improved and above all, more reasonably priced components for the PEMFC and the reforming of methanol is to work out the important technical basis for the subsequent economic conversion into a vehicle drive. (orig.) [Deutsch] Polymer-Elektrolyt-Membran-Brennstoffzellen (PEMFC) zeichnen sich durch hohe spezifische Leistungsdichten, wie sie im Bereich elektrischer Antriebe notwendig sind, hohe Wirkungsgrade und treibstoffabhaengig keine bzw. sehr niedrige Schadstoffemission aus. Im Rahmen eines BMBF-gefoerderten Leitprojekts sollen mit der Entwicklung verbesserter, vor allem kostenguenstiger Komponenten fuer die PEMFC und die Methanol-Reformierung wesentliche technologische Grundlagen fuer die anschliessende wirtschaftliche Umsetzung als Fahrzeugantrieb erarbeitet werden. (orig.)

  7. A comparison of radiation doses and risks between spent fuel transport/storage and selected non-nuclear activities

    International Nuclear Information System (INIS)

    Pennington, C.W.

    2003-01-01

    Spent fuel transport and storage have achieved an exemplary safety record over four decades within both the United States (US) and the global community at large. This paper offers an assessment demonstrating the safety of spent fuel transport and storage packagings relative to currently accepted but unregulated non-nuclear activities and practices within society. Over the last quarter of a century, several spent fuel transport and storage packaging test programmes have produced data that allow calculation of potential releases and population doses resulting from a terrorist attack. The US Department of Energy (DOE) has used this information to develop projected worst-case, low probability population exposures as part of the Final Environmental Impact Statement (FEIS) for the Yucca Mountain repository. The paper discusses potential population exposures from these packagings based on analysis and testing under beyond-design-basis (BDB) events, including missile attacks, and then defines and defends an acceptance criterion for the bounding outcomes of these events, based upon current accepted activities within society that produce high radiation doses to the general public. These activities involve unregulated technologies and practices within society that yield population doses significantly exceeding those that would result from such hypothetical and highly improbable events as a terrorist missile attack on a spent fuel transport or storage packaging. In particular, technologically enhanced natural radiation (TENR) exposures from building materials, farming, and masonry construction are highlighted. Recent landmark work by the US National Academy of Sciences (NAS) and by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) are cited in support of this assessment, along with work from the US Environmental Protection Agency (EPA). From this compelling evidence, it is concluded that spent fuel transport and storage represent a low

  8. Massachusetts Fuel Cell Bus Project: Demonstrating a Total Transit Solution for Fuel Cell Electric Buses in Boston

    Energy Technology Data Exchange (ETDEWEB)

    2017-05-22

    The Federal Transit Administration's National Fuel Cell Bus Program focuses on developing commercially viable fuel cell bus technologies. Nuvera is leading the Massachusetts Fuel Cell Bus project to demonstrate a complete transit solution for fuel cell electric buses that includes one bus and an on-site hydrogen generation station for the Massachusetts Bay Transportation Authority (MBTA). A team consisting of ElDorado National, BAE Systems, and Ballard Power Systems built the fuel cell electric bus, and Nuvera is providing its PowerTap on-site hydrogen generator to provide fuel for the bus.

  9. Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1985-05-01

    Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

  10. Spent nuclear fuels project characterization data quality objectives strategy

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Thornton, T.A.; Redus, K.S.

    1994-12-01

    A strategy is presented for implementation of the Data Quality Objectives (DQO) process to the Spent Nuclear Fuels Project (SNFP) characterization activities. Westinghouse Hanford Company (WHC) and the Pacific Northwest Laboratory (PNL) are teaming in the characterization of the SNF on the Hanford Site and are committed to the DQO process outlined in this strategy. The SNFP characterization activities will collect and evaluate the required data to support project initiatives and decisions related to interim safe storage and the path forward for disposal. The DQO process is the basis for the activity specific SNF characterization requirements, termed the SNF Characterization DQO for that specific activity, which will be issued by the WHC or PNL organization responsible for the specific activity. The Characterization Plan prepared by PNL defines safety, remediation, and disposal issues. The ongoing Defense Nuclear Facility Safety Board (DNFSB) requirement and plans and the fuel storage and disposition options studies provide the need and direction for the activity specific DQO process. The hierarchy of characterization and DQO related documentation requirements is presented in this strategy. The management of the DQO process and the means of documenting the DQO process are described as well as the tailoring of the DQO process to the specific need of the SNFP characterization activities. This strategy will assure stakeholder and project management that the proper data was collected and evaluated to support programmatic decisions

  11. Externalities of fuel cycles 'ExternE' project. Summary report

    International Nuclear Information System (INIS)

    Holland, M.; Berry, J.

    1994-01-01

    There is a growing requirement for policy analysts to take account of the environment in their decision making and to undertake the specified cost-benefit analysis. Within the European Union this is reflected in the 5th Environmental Action Programme, and the Commission's White Paper entitled 'Growth, competitiveness, employment and the ways forward to the 21st century'. This has led to a need for evaluation of environmental externalities. The ExternE Project commenced in 1991 as the European part of a collaborative study between the European Commission and the US Department of Energy. It aims to be the first systematic approach to the evaluation of external costs of a wide range of different fuel cycles. The project will result in an operational accounting framework for the quantification and monetarisation of priority environmental and other externalities. This framework will allow the calculation of the marginal external costs and benefits for specific power plants, at specific sites using specified technologies. There are three major phases in the project. Phase I was undertaken in collaboration with the US Department of Energy. In this phase the teams jointly developed the conceptual approach and methodology and shared scientific information for application to a number of fuel cycles. On the European side work concentrated on the nuclear and coal fuel cycles which together were expected to raise many of the fundamental issues in fuel cycle analysis. The project is currently nearing completion of Phase 2. During this phase the methodology has been applied to a wide range of different fossil, nuclear and renewable fuel cycles for power generation and energy conservation options. Also a series of National Implementation Programmes are underway in which the methodology and accounting framework are being applied to reference sites throughout Europe. In addition the general methodology is being extended to address the evaluation of externalities associated with

  12. Thermal hydraulic feasibility assessment of the spent nuclear fuel project

    International Nuclear Information System (INIS)

    Heard, F.J.

    1996-01-01

    A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The goal was to develop a series of thermal-hydraulic models that could respond to all process and safety related issues that may arise pertaining to the SNFP, as well as provide a basis for validation of the results. Results show that there is a reasonable envelope for process conditions and requirements that are thermally and hydraulically acceptable

  13. Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy

    International Nuclear Information System (INIS)

    TRAWINSKI, B.J.

    2000-01-01

    The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system

  14. The reprocessing of fast reactor fuels - the TOR project

    International Nuclear Information System (INIS)

    Calame-Longjean, A.; Le Bouhellec, J.; Schwob, Y.

    1982-01-01

    A description is given of development work on the proposed new French facility for the reprocessing of fast reactor fuel. This is the TOR facility (Traitement des Oxydes Rapides). Block diagrams give details of the TOR project as a whole and of the main line and R and D line of the TOR 1 facility which is a new works devoted to the head of the process. Modifications to existing plant which will form the TOR 2 and TOR 3 facilities are also described. (U.K.)

  15. Analysis of radiation doses for a transportation system and its interface operations for commercial spent fuel

    International Nuclear Information System (INIS)

    Schneider, K.J.; Ross, W.A.; Smith, R.I.; Wilmot, E.L.

    1987-07-01

    This paper gives the results of estimates of aggregated radiation doses to the affected public and workers in the US that would be associated with loading spent fuel at the reactors, transporting the spent fuel by truck and rail, and receiving and unloading the spent fuel at a deep geological repository. The estimates are for a postulated transportation-related system using current state-of-the-art technology, if employed in the high-level waste management system in the future, and the approximate dose reduction from some potential system improvements. The results of the study provide a starting point for the US Department of Energy (DOE) to develop an improved transportation system that is cost effective, safe, and results in low radiation doses. 4 refs., 1 figs., 5 tabs

  16. Dose assessment for public at the hypothetical submergence of a fresh MOX fuel package

    International Nuclear Information System (INIS)

    Tsumune, Daisuke; Saegusa, Toshiari; Suzuki, Hiroshi; Maruyama, Koki

    2000-01-01

    For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel is transported safety on the sea. To gain the public acceptance for this transport, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unknown reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

  17. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

    International Nuclear Information System (INIS)

    Chandler, M.C.; Ketusky, E.T.; Thoman, D.C.

    1995-01-01

    Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ''representative'' set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions

  18. 14 CFR 26.37 - Pending type certification projects: Fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Pending type certification projects: Fuel tank flammability. 26.37 Section 26.37 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... AIRPLANES Fuel Tank Flammability § 26.37 Pending type certification projects: Fuel tank flammability. (a...

  19. NRC source term assessment for incident response dose projections

    International Nuclear Information System (INIS)

    Easley, P.; Pasedag, W.

    1984-01-01

    The NRC provides advice and assistance to licensees and State and local authorities in responding to accidents. The TACT code supports this function by providing source term projections for two situations during early (15 to 60 minutes) accident response: (1) Core/containment damage is indicated, but there are no measured releases. Quantification of a predicted release permits emergency response before people are exposed. With TACT, response personnel can estimate releases based on fuel and cladding conditions, coolant boundary and containment integrity, and mitigative systems operability. For this type of estimate, TACT is intermediate between default assumptions and time-consuming mechanistic codes. (2) A combination of plant status and limited release data are available. For this situation, iterations between predictions based on known conditions which are compared to measured releases gives reasonable confidence in supplemental source term information otherwise unavailable: nuclide mix, releases not monitored, and trending or abrupt changes. The assumptions and models used in TACT, and examples of its use, are given in this paper

  20. Fuel transfer system ALARA design review - Project A.15

    International Nuclear Information System (INIS)

    KUEBERTH, L.R.

    2001-01-01

    One mission of the Spent Nuclear Fuel (SNF) Project is to move the SNF from the K Basins in the Hanford 100K Area to an interim dry storage at the Canister Storage Building (CSB) in the Hanford 200 East Area. The Fuel Transfer System (FTS) is a subproject that will move the SNF from the 105K East (KE) Facility to the 105K West (KW) Facility. The SNF will be treated for shipment to the Cold Vacuum Drying (CVD) facility at the KW Basin. The SNF canisters will be loaded underwater into a Shielded Transfer Cask (STC) in the KE Basin. The fully loaded STC will be brought out of the water and placed into a Cask Transfer Overpack (CTO) by the STC Straddle Carrier. As the STC is removed from the water, it will be washed down with demineralized water by an manual rinse system. The CTO with the STC inside will be placed on a transport trailer and transferred to the KW Basin as an intra-facility transfer. The CTO will be unloaded from the shipping trailer at the KW Basin and the STC will be removed from the CTO. The STC will then be lowered into the KW Basin water and the fuel will be removed. The SNF will then be processed for shipment to the CVD. As soon as all of the fuel has been removed from the STC, the cask will be removed from the KW Basin water and placed into the CTO. The CTO will again be placed on the trailer for transport back to the KE Basin where the entire cycle will be repeated approximately 400 times. This document records the As Low As Reasonably Achievable (ALARA) findings and design recommendations/requirements by the SNF Project noted during the Final Design Review of the STC, CTO, STC Transfer System, Annexes and Roadways for support of FTS. This document is structured so that all statements that include the word ''shall'' represent design features that have been or will be implemented within the project scope. Statements that include the words ''should'' or ''recommend'' represent ALARA design features to be evaluated for future implementation

  1. Prediction of dose and field mapping around a shielded plutonium fuel fabrication glovebox

    International Nuclear Information System (INIS)

    Strode, J.N.; Soldat, K.L.; Brackenbush, L.W.

    1984-01-01

    Westinghouse Hanford Company, as the Department of Energy's (DOE) prime contractor for the operation of the Hanford Engineering Development Laboratory (HEDL), is responsible for the development of the Secure Automated Fabrication (SAF) Line which is to be installed in the recently constructed Fuels and Materials Examination Facility (FMEF). The SAF Line will fabricate mixed-oxide (MOX) fuel pins for the Fast Flux Test Facility (FFTF) at an annual throughput rate of six (6) metric tons (MT) of MOX. The SAF Line will also demonstrate the automated manufacture of fuel pins on a production-scale. This paper describes some of the techniques used to reduce personnel exposure on the SAF Line, as well as the prediction and field mapping of doses from a shielded fuel fabrication glovebox. Tables are also presented from which exposure rate estimates can be made for plutonium recovered from fuels having different isotopic compositions as a result of varied burnup

  2. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

    Mourogov, V. M.; Juhn, P. E.

    2003-01-01

    In response to two IAEA General Conference Resolutions in September 2000, the IAEA has launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) in May 2001. As of February 2003, 12 IAEA Member States and the European Commission have become members of INPRO. In total, 19 cost-free experts have been nominated by these Member States and the European Commission to work for the INPRO project at the IAEA. Four meetings of the INPRO Steering Committee (SC), which is the decision and review body of INPRO, were held, two in 2001 and another two in 2002. The objective of INPRO, which is composed of two phases (Phase 1 and Phase 2), is to support safe, economic and proliferation resistant use of nuclear technology, in a sustainable manner, to meet the global energy needs in the next 50 years and beyond. During Phase 1, work is also subdivided in two sub phases: The currently on-going Phase 1A is focussing on the selection of criteria and development of methodologies and guidelines for the comparison of different reactor and fuel cycle concepts and approaches, taking into account the compilation and review of such concepts and approaches, and determination of user requirements in the areas of economics; environment; safety; proliferation-resistance; and cross cutting issues. The preliminary results of Phase 1A with respect to user requirements are summarized in the paper

  3. Parameter calculation tool for the application of radiological dose projection codes

    International Nuclear Information System (INIS)

    Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J.; Tijerina S, F.

    2016-09-01

    The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)

  4. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    Science.gov (United States)

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  5. Calculation of radiation dose rates from a spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask

  6. Hanford Environmental Dose Reconstruction Project, Quarterly report, September--November 1993

    International Nuclear Information System (INIS)

    Cannon, S.D.; Finch, S.M.

    1993-01-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation doses that individuals and populations could have received from nuclear operations at Hanford since 1944. The project is divided into the following technical tasks. These tasks correspond to the path radionuclides followed from release to impact on humans (dose estimates); Source Terms, Environmental Transport, Environmental Monitoring Data, Demography, Food Consumption, and Agriculture, and Environmental Pathways and Dose Estimates

  7. KWIKPLAN: a computer program for projecting the annual requirements of nuclear fuel cycle operations

    International Nuclear Information System (INIS)

    Salmon, R.; Kee, C.W.

    1977-06-01

    The computer code KWIKPLAN was written to facilitate the calculation of projected nuclear fuel cycle activities. Using given projections of power generation, the code calculates annual requirements for fuel fabrication, fuel reprocessing, uranium mining, and plutonium use and production. The code uses installed capacity projections and mass flow data for six types of reactors to calculate projected fuel cycle activities and inventories. It calculates fissile uranium and plutonium flows and inventories after allowing for an economy with limited reprocessing capacity and a backlog of unreprocessed fuel. All calculations are made on a quarterly basis; printed and punched output of the projected fuel cycle activities are made on an annual basis. Since the punched information is used in another code to determine waste inventories, the code punches a table from which the effective average burnup can be calculated for the fuel being reprocessed

  8. The IAEA International Project on Innovative Reactors and Fuel Systems

    International Nuclear Information System (INIS)

    Mourogov, V.M.

    2001-01-01

    Full text: Nuclear power is faced with a dilemma. From one side, there is no doubt (particularly in our community) that nuclear power can play an outstanding role in a sustainable energy system worldwide due to its well known potential advantages. From the other side we have near-term nuclear power projections and prospects that are not so promising. In 2000 nuclear's share was 3% of total global electricity capacity additions which is more then three times lower that nuclear's 10% share of today's currently installed global capacity. It is also unfortunate that nuclear capacity additions in developing countries, where the main increase in energy demand is expected, are relatively insignificant compared to fossil and hydro capacity additions in recent years. Most near-term projections show no drastic changes in these recent trends How can we address this dilemma? If the nuclear power sector is to increase its role, it cannot simply continue to do what it has been doing and expect that factors outside its control, such as fossil fuel prices or environmental taxes, will change to make nuclear power's prospects more favorable. To reach a different outcome than that indicated by current near- and intermediate-term trends, something must be done within the nuclear community to generate new technological solutions. The challenge is to look to the future, to identify what innovations and new directions - that build upon and make good use of existing expertise and accomplishments - are most promising for helping nuclear power capture a growing share of a growing market. There are several challenges that we have to deal to facilitate large-scale global nuclear power development. These are: achieving economic competitiveness of new NPPs in most parts of the world; successfully demonstrating effective nuclear waste management; responsiveness to public safety concerns; responsiveness to proliferation concerns. And as a result building support for nuclear power among the public

  9. Estimation of gamma dose rate from hulls and shield design for the hull transport cask of Fuel Reprocessing Plant (FRP)

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    In Fuel Reprocessing Plant (FRP), un-dissolved clad of fuel pins known as hulls are the major sources of high level solid waste. Safe handling, transport and disposal require the estimation of radioactivity as a consequent of gamma dose rate from hulls in fast reactor fuel reprocessing plant in comparison with thermal reactor fuel. Due to long irradiation time and low cooling of spent fuel, the evolution of activation products 51 Cr, 58 Co, 54 Mn and 59 Fe present as impurities in the fuel clad are the major sources of gamma radiation. Gamma dose rate from hull container with hulls from Fuel Sub Assembly (FSA) and Radial Sub Assembly (RSA) of Fuel Reprocessing Plant (FRP) was estimated in order to design the hull transport cask. Shielding computations were done using point kernel code, IGSHIELD. This paper describes the details of source terms, estimation of dose rate and shielding design of hull transport cask in detail. (author)

  10. FY 1992 revised task plans for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Shipler, D.B.

    1992-04-01

    The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses from Hanford Site operations since 1944 to populations and individuals. The primary objectives of work to be performed in FY 1992 is to determine the appropriate scope (space, time, and radionuclides, pathways and individuals/population groups) and accuracy (level of uncertainty in dose estimates) for the project. Another objective is to use a refined computer model to estimate Native American tribal doses and individual doses for the Hanford Thyroid Disease Study (HTDS). Project scope and accuracy requirements defined in FY 1992 can translated into model and data requirements that must be satisfied during FY 1993

  11. From waste to traffic fuel -projects. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rasi, S; Lehtonen, E; Aro-Heinilae, E [and others

    2012-11-01

    The main objective of the project was to promote biogas production and its use as traffic fuel. The aims in the four Finnish and two Estonian case regions were to reduce the amount and improve the sustainable use of waste and sludge, to promote biogas production, to start biogas use as traffic fuel and to provide tools for implementing the aims. The results of this study show that achieving the food waste prevention target will decrease greenhouse gas emissions by 415 000 CO{sub 2}-eq tons and result in monetary savings for the waste generators amounting to almost 300 euro/ capita on average in all case regions in 2020. The results show that waste prevention should be the first priority in waste management and the use of waste materials as feedstock for energy production the second priority. In total 3 TWh energy could be produced from available biomass in the studied case regions. This corresponds to the fuel consumption of about 300 000 passenger cars. When a Geographical Information System (GIS) was used to identify suitable biogas plant site locations with particular respect to the spatial distribution of available biomass, it was found that a total of 50 biogas plants with capacity varying from 2.1 to 14.5 MW could be built in the case regions. This corresponds to 2.2 TWh energy and covers from 5 to 40% of the passenger car fuel consumption in these regions. Using all produced biogas (2.2 TWh energy) for vehicle fuel GHG emissions would lead to a 450 000 t CO{sub 2}-eq reduction. The same effect on emissions would be gained if more than 100 000 passenger cars were to be taken off the roads. On average, the energy consumed by biogas plants represents approximately 20% of the produced energy. The results also show that biomethane production from waste materials is profitable. In some cases the biomethane production costs can be covered with the gained gate fees. The cost of biomethane production from agricultural materials is less than 96 euro/MWh{sub th

  12. Project Profile: Hydrogen Fuel Cell Mobile Lighting Tower (HFCML)

    Science.gov (United States)

    McLaughlin, Russell

    2013-01-01

    NASA is committed to finding innovative solutions that improve the operational performance of ground support equipment while providing environment and cost benefits, as well. Through the Hydrogen Fuel Cell Mobile Lighting Tower (HFCML) project, NASA gained operational exposure to a novel application of high efficiency technologies. Traditionally, outdoor lighting and auxiliary power at security gates, launch viewing sites, fallback areas, outage support, and special events is provided by diesel generators with metal halide lights. Diesel generators inherently contribute to C02, NOx, particulate emissions, and are very noisy. In 2010, engineers from NASA's Technology Evaluation for Environmental Risk Mitigation Principal Center (TEERM) introduced KSC operations to a novel technology for outdoor lighting needs. Developed by a team led by Sandia National Laboratory (SNL), the technology pairs a 5kW hydrogen fuel cell with robust high efficiency plasma lights in a towable trailer. Increased efficiency, in both the fuel cell power source and lighting load, yields longer run times between fueling operations while providing greater auxiliary power. Because of the unit's quiet operation and no exhaust fumes, it is capable of being used indoors and in emergency situations, and meets the needs of all other operational roles for metal halide/diesel generators. The only discharge is some water and warm air. Environmental benefits include elimination of diesel particulate emissions and estimated 73% greenhouse gas emissions savings when the hydrogen source is natural gas (per GREET model). As the technology matures the costs could become competitive for the fuel cell units which are approximately 5 times diesel units. Initial operational . concerns included the hydrogen storage tanks and valves, lightning safety/grounding, and required operating and refueling procedures. TEERM facilitated technical information exchange (design drawings, technical standards, and operations

  13. Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

    International Nuclear Information System (INIS)

    Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

    2003-01-01

    One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

  14. Spent Nuclear Fuel (SNF) Project Design Verification and Validation Process

    International Nuclear Information System (INIS)

    OLGUIN, L.J.

    2000-01-01

    This document provides a description of design verification and validation activities implemented by the Spent Nuclear Fuel (SNF) Project. During the execution of early design verification, a management assessment (Bergman, 1999) and external assessments on configuration management (Augustenburg, 1999) and testing (Loscoe, 2000) were conducted and identified potential uncertainties in the verification process. This led the SNF Chief Engineer to implement corrective actions to improve process and design products. This included Design Verification Reports (DVRs) for each subproject, validation assessments for testing, and verification of the safety function of systems and components identified in the Safety Equipment List to ensure that the design outputs were compliant with the SNF Technical Requirements. Although some activities are still in progress, the results of the DVR and associated validation assessments indicate that Project requirements for design verification are being effectively implemented. These results have been documented in subproject-specific technical documents (Table 2). Identified punch-list items are being dispositioned by the Project. As these remaining items are closed, the technical reports (Table 2) will be revised and reissued to document the results of this work

  15. 18-months fuel cycle engineering and its project management of the Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    Fu Xiangang; Jiao Ping; Liu Yong; Wu Zhiming

    2002-01-01

    The author introduces aspects related to the performing of 18-months fuel cycle engineering evaluation to the Daya Bay nuclear power plant, including the assessment on proposed technical solutions, appointment to the contractors, breaking down and implementation of project, experience on the project management and risk control, and etc. And it also briefs the prompting to the localization of the long fuel cycle engineering technology and AFA 3G fuel manufacturing and design technology via adequate technology transferring of this project

  16. Application of dose evaluation of the MCNP code for interim spent fuel cask storage facility

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Iimoto, Takeshi; Ishikawa, Satoshi; Tsuboi, Takafumi; Teramura, Masahiro; Okamura, Tomomi; Narumiya, Yoshiyuki

    2007-01-01

    The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results. (author)

  17. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy's Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses

  18. Training implementation matrix. Spent Nuclear Fuel Project (SNFP)

    International Nuclear Information System (INIS)

    EATON, G.L.

    2000-01-01

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently

  19. Spent Nuclear Fuel Project Cold Vacuum Drying Facility Operations Manual

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1999-01-01

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, (Cold Vacuum Drying Facility Design Requirements), Rev. 4. and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  20. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  1. Development of a non-engine fuel injector deposit test for alternative fuels (ENIAK-project)

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Hajo; Pohland vom Schloss, Heide [OWI - Oel Waerme Institut GmbH, Herzogenrath (Germany)

    2013-06-01

    Deposit formation in and on the injectors of diesel engines may lead to injector malfunction, resulting in a loss in power, rough engine operation and poor emission levels. Poor Biodiesel quality, contamination with copper and zinc as well as undesired reactions between (several) additives and biodiesel components are known causes for nozzle fouling. Therefore, good housekeeping when using biodiesel is required, and all additives have to pass a no-harm test concerning injector fouling. The standard fouling tests are two engine tests: The XUD9-test (CEC F-23-01) and the DW-10-test (CEC DF 98-08). The XUD9 is a cost efficient, fast and proven testing method. It uses, however, an obsolete indirect injection diesel engine and cannot reproduce internal diesel injector deposits (IDID). The newer DW10 test is complex, costly and designed for high stress. This reduces the engine life and leads to a fuel consumption of approximately 1,000 1 per test, both contributing to the high costs of the test. The ENIAK-Project is funded by the FNR (''Fachagentur Nachwachsende Rohstoffe'', Agency for Renewable Resources) and conducted in cooperation with AGQM, ASG and ERC. Its main goal is the development, assembly, commissioning, and evaluation of a non-engine fuel injector test. It uses a complete common rail system. The injection takes place in a self-designed reactor instead of an engine, and the fuel is not combusted, but re-condensed and pumped in a circle, leading to a low amount of fuel required. If the test method proves to be as reliable as expected, it can be used as an alternative test method for injector fouling with low requirements regarding infrastructure on the testing site and sample volume. (orig.)

  2. Dose rate analysis for Tank 101 AZ (Project W151)

    International Nuclear Information System (INIS)

    Schwarz, R.A.; Hillesland, K.E.; Carter, L.L.

    1994-11-01

    This document describes the expected dose rates for modification to tank 101 AZ including modifications to the steam coil, mixer pump, and temperature probes. The thrust of the effort is to determine dose rates from: modification of a steam coil and caisson; the installation of mixer pumps; the installation of temperature probes; and estimates of dose rates that will be encountered while making these changes. Because the dose rates for all of these configurations depend upon the photon source within the supernate and sludge, comparisons were also made between measured dose rates within a drywell and the corresponding calculated dose rates. The calculational tool used is a Monte Carlo (MCNP 2 ) code since complicated three dimensional geometries are involved. A summary of the most important results of the entire study is given in Section 2. The basic calculational geometry model of the tank is discussed in Section 3, along with a tabulation of the photon sources that were used within the supernate and the sludge, and a discussion of uncertainties. The calculated dose rates around the steam coil and caisson before and after modification are discussed in Section 4. The configuration for the installation of the mixer pumps and the resulting dose rates are given in Section 5. The predicted changes in dose rates due to a possible dilution of the supernate source are given in Section 6. The calculational configuration used to model the installation of temperature probes and the resulting predicted dose rates are discussed in Section 7. Finally, comparisons of measured to calculated dose rates within a drywell are summarized in Section 8. Extended discussions of calculational models and Monte Carlo optimization techniques used are included in Appendix A

  3. Spent nuclear fuel project high-level information management plan

    Energy Technology Data Exchange (ETDEWEB)

    Main, G.C.

    1996-09-13

    This document presents the results of the Spent Nuclear Fuel Project (SNFP) Information Management Planning Project (IMPP), a short-term project that identified information management (IM) issues and opportunities within the SNFP and outlined a high-level plan to address them. This high-level plan for the SNMFP IM focuses on specific examples from within the SNFP. The plan`s recommendations can be characterized in several ways. Some recommendations address specific challenges that the SNFP faces. Others form the basis for making smooth transitions in several important IM areas. Still others identify areas where further study and planning are indicated. The team`s knowledge of developments in the IM industry and at the Hanford Site were crucial in deciding where to recommend that the SNFP act and where they should wait for Site plans to be made. Because of the fast pace of the SNFP and demands on SNFP staff, input and interaction were primarily between the IMPP team and members of the SNFP Information Management Steering Committee (IMSC). Key input to the IMPP came from a workshop where IMSC members and their delegates developed a set of draft IM principles. These principles, described in Section 2, became the foundation for the recommendations found in the transition plan outlined in Section 5. Availability of SNFP staff was limited, so project documents were used as a basis for much of the work. The team, realizing that the status of the project and the environment are continually changing, tried to keep abreast of major developments since those documents were generated. To the extent possible, the information contained in this document is current as of the end of fiscal year (FY) 1995. Programs and organizations on the Hanford Site as a whole are trying to maximize their return on IM investments. They are coordinating IM activities and trying to leverage existing capabilities. However, the SNFP cannot just rely on Sitewide activities to meet its IM requirements

  4. Radionuclide mass inventory, activity, decay heat, and dose rate parametric data for TRIGA spent nuclear fuels

    International Nuclear Information System (INIS)

    Sterbentz, J.W.

    1997-03-01

    Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) Low-enrichment Fuel Life Improvement Program (FLIP-LEU-1). Parametric activity data are tabulated for 145 important radionuclides that can be used to generate gamma-ray emission source terms or provide mass quantity estimates as a function of decay time. Fuel element decay heats and dose rates are also presented parametrically as a function of burnup and decay time. Dose rates are given at the fuel element midplane for contact, 3.0-feet, and 3.0-meter detector locations in air. The data herein are estimates based on specially derived Beginning-of-Life (BOL) neutron cross sections using geometrically-explicit TRIGA reactor core models. The calculated parametric data should represent good estimates relative to actual values, although no experimental data were available for direct comparison and validation. However, because the cross sections were not updated as a function of burnup, the actinide concentrations may deviate from the actual values at the higher burnups

  5. Radionuclide releases to the atmosphere from Hanford Operations, 1944--1972. Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.

    1994-05-01

    The purpose of the Hanford Environmental Dose Reconstruction Project is to estimate the radiation dose that individuals could have received as a result of radionuclide emissions since 1944 from the Hanford Site. The first step in determining dose is to estimate the amount and timing of radionuclide releases to air and water. This report provides the air release information.

  6. 77 FR 33158 - Plumas National Forest, California, Sugarloaf Hazardous Fuels Reduction Project

    Science.gov (United States)

    2012-06-05

    ... Fuels Reduction Project AGENCY: Forest Service, USDA. ACTION: Notice of intent to prepare an... National Forest (PNF) will prepare an environmental impact statement (EIS) on the Sugarloaf Hazardous Fuels... to the economic stability of rural communities through: fuels treatments; group selections (GS); area...

  7. 75 FR 16422 - Plumas National Forest, California, Keddie Ridge Hazardous Fuels Reduction Project

    Science.gov (United States)

    2010-04-01

    ... Hazardous Fuels Reduction Project AGENCY: Forest Service, USDA. ACTION: Notice of intent to prepare an... District will prepare and environmental impact statement (EIS) on the Keddie Ridge Hazardous Fuels...: fuels treatments, group selections, road improvements, and herbicide and mechanical applications in the...

  8. Factors affecting radiation doses from dedicated rail transport of spent reactor fuel

    International Nuclear Information System (INIS)

    Martin, J.E.

    1988-01-01

    This paper reports there are two exposure control concerns associated with the shipment of spent reactor fuel in dedicated trains -- compliance with transportation regulations for maximum allowable radiation levels, and minimizing the dose received by the general public. This article examines the methods used to calculate the dose equivalent rates alongside stationary (transport regulations) and moving trains (public exposure) of various lengths. The factors examined include the source term, the effect of overlapping radiation fields, the speed of the train, and the location of the population relative to the train. Trains made up of series of cars that individually meet transport regulations can, as a whole, exceed transport vehicle dose equivalent rate limits by up to 23% due to overlapping radiation fields. For moving trains and the worst case analyzed -- a person located 20 feet from the tracks and a train speed of 5 mph --- 141 rail cars would have to pass by to deliver a dose equivalent of 1 mrem

  9. Solar fuels production as a sustainable alternative for substituting fossil fuels: COSOLπ project

    Science.gov (United States)

    Hernando Romero-Paredes, R.; Alvarado-Gil, Juan José; Arancibia-Bulnes, Camilo Alberto; Ramos-Sánchez, Víctor Hugo; Villafán-Vidales, Heidi Isabel; Espinosa-Paredes, Gilberto; Abanades, Stéphane

    2017-06-01

    This article presents, in summary form, the characteristics of COSOLπ development project and some of the results obtained to date. The benefits of the work of this project will include the generation of a not polluting transportable energy feedstock from a free, abundant and available primary energy source, in an efficient method with no greenhouse gas emission. This will help to ensure energy surety to a future transportation/energy infrastructure, without any fuel import. Further technological development of thermochemical production of clean fuels, together with solar reactors and also with the possibility of determining the optical and thermal properties of the materials involved a milestone in the search for new processes for industrialization. With the above in mind, important national academic institutions: UAM, UNAM, CINVESTAV, UACH, UNISON among others, have been promoting research in solar energy technologies. The Goals and objectives are to conduct research and technological development driving high-temperature thermochemical processes using concentrated solar radiation as thermal energy source for the future sustainable development of industrial processes. It focuses on the production of clean fuels such as H2, syngas, biofuels, without excluding the re-value of materials used in the industry. This project conducts theoretical and experimental studies for the identification, characterization, and optimization of the most promising thermochemical cycles, and for the thorough investigation of the reactive chemical systems. It applies material science and nano-engineering to improve chemicals properties and stability upon cycling. The characterization of materials will serve to measure the chemical composition and purity (MOX fraction-1) of each of the samples. The characterizations also focus on the solid particle morphology (shape, size, state of aggregation, homogeneity, specific surface) images obtained from SEM / TEM and BET measurements. Likewise

  10. International project on innovative nuclear reactors and fuel cycles (INPRO)

    International Nuclear Information System (INIS)

    Omoto, A.

    2006-01-01

    The IAEA's project INPRO was initiated in order to provide a forum for discussion of experts and policy makers on all aspects of nuclear energy planning as well as on the development and deployment of innovative nuclear energy systems (INS). It brings together technology holders users and potential users to consider jointly the international and national actions required for achieving desired innovations in nuclear reactors and fuel cycles, but it pays particular attention to the needs of developing countries. Currently INPRO members count 24 including even three countries, which are not yet operating nuclear reactors. Its initial phase has produced an outlook into the future of the energy markets and defined basic principles, user requirements and criteria in the following areas as TECDOC1362 in June 2003; Economics, Environment, Fuel Cycle and Waste, Safety, Proliferation Resistance and Crosscutting Issues. This assessment methodology can be applied for screening an INS, comparing different INS to find a preferred INS consistent with the needs of a given state, and identifying RD and D needs. The methodology has be validated through case studies and updated as TECDOC1434 in December 2004. Currently, besides producing a manual for each chapter of TECDOC1434, six assessment studies of various INS options are being carried out and the number of such studies is increasing. Further several tasks are ongoing including modeling and analysis of global and regional balance of resources and INS deployment scenarios in order to gain the better perspective of future implication of INS deployment as well as to identify challenges and opportunities of INS. It is envisioned that INPRO will continue to develop with three planned major pillars of activity; methodology, infrastructure and coordination for planning of R and D activities. The paper discusses the progress and status of INPRO as well as the future prospect of INPRO activities

  11. Transient response and radiation dose estimates for breaches to a spent fuel processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Solbrig, Charles W., E-mail: soltechco@aol.com; Pope, Chad; Andrus, Jason

    2014-08-15

    Highlights: • We model doses received from a nuclear fuel facility from boundary leaks due to an earthquake. • The supplemental exhaust system (SES) starts after breach causing air to be sucked into the cell. • Exposed metal fuel burns increasing pressure and release of radioactive contamination. • Facility releases are small and much less than the limits showing costly refits are unnecessary. • The method presented can be used in other nuclear fuel processing facilities. - Abstract: This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

  12. Transient response and radiation dose estimates for breaches to a spent fuel processing facility

    International Nuclear Information System (INIS)

    Solbrig, Charles W.; Pope, Chad; Andrus, Jason

    2014-01-01

    Highlights: • We model doses received from a nuclear fuel facility from boundary leaks due to an earthquake. • The supplemental exhaust system (SES) starts after breach causing air to be sucked into the cell. • Exposed metal fuel burns increasing pressure and release of radioactive contamination. • Facility releases are small and much less than the limits showing costly refits are unnecessary. • The method presented can be used in other nuclear fuel processing facilities. - Abstract: This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities

  13. Grand Junction Projects Office Remedial Action Project Building 2 public dose evaluation. Final report

    International Nuclear Information System (INIS)

    Morris, R.

    1996-05-01

    Building 2 on the U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) site, which is operated by Rust Geotech, is part of the GJPO Remedial Action Program. This report describes measurements and modeling efforts to evaluate the radiation dose to members of the public who might someday occupy or tear down Building 2. The assessment of future doses to those occupying or demolishing Building 2 is based on assumptions about future uses of the building, measured data when available, and predictive modeling when necessary. Future use of the building is likely to be as an office facility. The DOE sponsored program, RESRAD-BUILD, Version. 1.5 was chosen for the modeling tool. Releasing the building for unrestricted use instead of demolishing it now could save a substantial amount of money compared with the baseline cost estimate because the site telecommunications system, housed in Building 2, would not be disabled and replaced. The information developed in this analysis may be used as part of an as low as reasonably achievable (ALARA) cost/benefit determination regarding disposition of Building 2

  14. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H p (10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces. (authors)

  15. Concept and development status of fast breeder reactor fuels in the FaCT project

    International Nuclear Information System (INIS)

    Maeda, S.; Suzuki, M.; Kaito, T.; Tanaka, K.; Abe, T.

    2013-01-01

    The fuel development and the conceptual design study have been progressed in the first phase of the FaCT project in Japan. Significant outcomes of key technologies related to fuel design, fuel properties, core materials, fuel fabrication have been provided. The prospects of these technologies have been identified. After the Fukushima accident, the research and development for reducing the amount and toxic level of radioactive wastes will be promoted more than before. These outcomes will be reflected on the future development

  16. The risks of the Taiwan research reactor spent fuel project

    International Nuclear Information System (INIS)

    1991-06-01

    The proposed action is to transport up to 118 spent fuel rods, to include canned spent fuel rod particulates immobilized on filters, from a research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the Receiving Basin for Offsite Fuels at the Savannah River Site (SRS). At SRS, the spent fuel will be reprocessed to recover uranium and plutonium. 55 refs., 8 tabs

  17. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-02-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institute, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make a specification document for the fuel pellet, the sphere-pac fuel particles, the vipac fuel particles, and the fuel pin. JNC should make a fabrication drawing for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Netherlands). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Netherlands). This fabrication drawing has been made under the design assignment with PSI, and consists of the drawing of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications has been discussed and agreed among all partners. In this report, the revised fabrication drawings will be shown. Based on the commission of Plutonium Fuel Technology Group, Advanced Fuel Recycle Technology Division, this design work has been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center. (author)

  18. Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version 2

    International Nuclear Information System (INIS)

    Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

    2002-10-01

    Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institut, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make specification documents for the fuel pellet, the sphere-pac fuel particles, the vipac fuel fragments, and the fuel segment fabrication. JNC should make the fabrication drawings for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Holland). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Holland). The fabrication drawings have been made under the design assignment with PSI, and consist of the drawings of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications was discussed and agreed among all partners. According to this agreement, the fabrication drawings were revised in January 2002. After the earlier revision, the shape of particle retainer to be made by PSI was modified from its drawing beforehand delivered. In this report, the fabrication drawings revised again will be shown, and the fabrication procedure (welding Qualification Tests) will be modified in accordance with the result of discussion on the 3rd technical meeting held in September 2002. These design works have been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center under the commission of Plutonium Fuel

  19. Hanford Spent Nuclear Fuel Project: Recommended path forward. Volume 2: Alternatives and path forward evaluation

    International Nuclear Information System (INIS)

    Fulton, J.C.

    1994-10-01

    The Hanford Spent Nuclear Fuel Project has completed an evaluation of four alternatives for expediting the removal of spent nuclear fuel from the K Basins and stabilizing and placing the fuel into interim storage. Four alternatives were compared: (1) Containerizing fuel in the K Basins, transporting fuel to a facility for stabilization, and interim storage of stabilized fuel in a dry storage facility (DSF); (2) Containerizing fuel in the K Basins, transporting fuel to a wet temporary staging facility, moving fuel to a facility for stabilization, and transporting stabilized fuel to an interim DSF; (3) Containerizing fuel in the K Basins in multi-canister overpacks, transporting fuel directly to a stabilization facility for passivation in the overpack, and interim storage of stabilized fuel in a DSF; (4) Packaging fuel for transport overseas and shipping fuel to a foreign reprocessing facility for reprocessing with eventual return of U, Pu and vitrified high level waste. The comparative evaluation consisted of a multi-attribute utility decision analysis, a public, worker and environmental health risk assessment, and a programmatic risk evaluation. The evaluation concluded that the best Path Forward combines the following concepts: Removal of K Basin fuel and sludge is uncoupled from the operation of a stabilization facility; A storage capability is provided to act as a lag storage or staging operation for overpack fuel containers as they are removed from the K Basins; Metal fuel drying and passivation should be maintained as the fuel stabilization process with the option of further refinements as more information becomes available; and The near term NEPA strategy should focus on expeditious removal of fuel and sludge from K Basins and placing overpacked fuel in temporary storage

  20. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  1. Methods and calculations for regional, continental, and global dose assessments from a hypothetical fuel reprocessing facility

    International Nuclear Information System (INIS)

    Schubert, J.F.; Kern, C.D.; Cooper, R.E.; Watts, J.R.

    1978-01-01

    The Savannah River Laboratory (SRL) is coordinating an interlaboratory effort to provide, test, and use state-of-the-art methods for calculating the environmental impact to an offsite population from the normal releases of radionuclides during the routine operation of a fuel-reprocessing plant. Results of this effort are the estimated doses to regional, continental, and global populations. Estimates are based upon operation of a hypothetical reprocessing plant at a site in the southeastern United States. The hypothetical plant will reprocess fuel used at a burn rate of 30 megawatts/metric ton and a burnup of 33,000 megawatt days/metric ton. All fuel will have been cooled for at least 365 days. The plant will have a 10 metric ton/day capacity and an assumed 3000 metric ton/year (82 percent online plant operation) output. Lifetime of the plant is assumed to be 40 years

  2. Estimated effects on radiation doses from alternatives in a spent fuel transportation system

    International Nuclear Information System (INIS)

    Schneider, K.J.; Ross, W.A.; Smith, R.I.

    1988-07-01

    This paper contains the results of a study of estimated radiation doses to the public and workers from the transport of spent fuel from commercial nuclear power reactors to a geologic repository. A postulated reference rail/legal-weight truck transportation system is defined that would use current transportation technology, and provide a breakdown of activities and time/distance/dose-rate estimates for each activity within the system. Collective doses are estimated for each of the major activities at the reactor site, in transit, and at the repository receiving facility. Annual individual doses to the maximally exposed individuals or groups of individuals are also estimated. The dose-reduction potentials and costs are estimated for a total of 17 conceptual alternatives and subalternatives to the postulated reference system. Most of the alternatives evaluated are estimated to provide both cost and dose reductions. The major conclusion is that the potential exists for significant future reductions in radiation doses to the public and workers and for reductions in costs compared to those based on a continuation of past practices in the US

  3. Estimated effects on radiation doses from alternatives in a spent fuel transportation system

    International Nuclear Information System (INIS)

    Schneider, K.J.; Ross, W.A.; Smith, R.I.

    1988-01-01

    This paper contains the results of a study of estimated radiation doses to the public and workers from the transport of spent fuel from commercial nuclear power reactors to a geologic repository. A postulated reference rail/legal-weight truck transportation system is defined that would use current transportation technology, and provide a breakdown of activities and time/distance/dose-rate estimates for each activity within the system. Collective doses are estimated for each of the major activities at the reactor site, in transit, and at the repository receiving facility. Annual individual doses to the maximally exposed individuals or groups of individuals also estimated. The dose-reduction potentials and costs are estimated for a total of 17 conceptual alternatives and subalternatives to the postulated reference system. Most of the alternatives evaluated are estimated to provide both cost and dose reductions. The major conclusion is that the potential exists for significant future reductions in radiation doses to the public and workers and for reductions in costs compared to those based on a continuation of past practices in the U.S

  4. Study on the evaluation method of radiation dose rate around spent fuel shipping casks

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1986-01-01

    This study aims at developing a simple calculation method which can evaluate radiation dose rate around casks with high accuracy in a short time. The method is based on a concept of the radiation shielding characteristics of cask walls. The concept was introduced to replace for ordinary radiation shielding calculation which requires a long calculation time and a large memory capacity of a computer in the matrix calculation. For the purpose of verifying the accuracy and reliability of the new method, it was applied to the analysis of the dose rate distribution around actual casks, which had been measured. The results of the analysis revealed that the newly proposed method was excellent for the forecast of radiation dose rate distribution around casks in view of the accuracy and calculation time. The short calculation time and high accuracy by the proposed method were attained by dividing the whole procedure of ordinary fine radiation shielding calculation into the calculation of radiation dose rate on a cask surface by the matrix expression of the characteristic function and the calculation of dose rate distribution using the simple analytical expression of dose rate distribution around casks. The effect of the heterogeneous array of spent fuel in different burnup state on dose rate distribution around casks was evaluated by this method. (Kako, I.)

  5. Correlation between scatter radiation dose at height of operator's eye and dose to patient for different angiographic projections

    International Nuclear Information System (INIS)

    Leyton, Fernando; Nogueira, Maria S.; Gubolino, Luiz A.; Pivetta, Makyson R.; Ubeda, Carlos

    2016-01-01

    Studies have reported cases of radiation-induced cataract among cardiology professionals. In view of the evidence of epidemiological studies, the ICRP recommends a new threshold for opacities and a new radiation dose to eye lens limit of 20 mSv per year for occupational exposure. The aim of this paper is to report scattered radiation doses at the height of the operator's eye in an interventional cardiology facility without considering radiation protection devices and to correlate these values with different angiographic projections and operational modes. Measurements were taken in a cardiac laboratory with an angiography X-ray system equipped with flat-panel detector. PMMA plates of 30×30×5 cm were used with a thickness of 20 cm. Measurements were taken in two fluoroscopy modes (low and normal, 15 pulses/s) and in cine mode (15 frames/s). Four angiographic projections were used: anterior posterior; lateral; left anterior oblique caudal (spider); and left anterior oblique cranial, with a cardiac protocol for patients weighing between 70 and 90 kg. Measurements of phantom entrance dose rate and scatter dose rate were performed with two Unfors Xi plus detectors. The detector measuring scatter radiation was positioned at the usual distance of the cardiologist's eyes during working conditions. There is a good linear correlation between the kerma area product and scatter dose at the lens. Experimental correlation factors of 2.3, 12.0, 12.2 and 17.6 μSv/Gy cm2 were found for different projections. PMMA entrance dose rates for low and medium fluoroscopy and cine modes were 13, 39 and 282 mGy/min, respectively, for AP projection. - Highlights: • A method is presented to estimate the scatter radiation dose at operator eye height. • The method allows estimating scatter radiation dose measuring ambient dose equivalent. • Operator could exceed threshold for lens opacities if protection tools are not used. • There is a good linear correlation between kerma

  6. A model of the dose rate calculation for a spent fuel storage structure by Monte Carlo method using the modulated code system SCALE 4.4a

    International Nuclear Information System (INIS)

    Pantazi, D.; Mateescu, S.; Stanciu, M.; Mete, M.

    2001-01-01

    The modulated code system SCALE is used to perform a standardized shielding analysis for any facility containing spent fuel: handling devices, transport cask, intermediate and final storage facility. The neutron and gamma sources as well as the dose rates can be obtained using either discrete-ordinates or Monte Carlo methods. The shielding analysis control modules (SAS1, SAS2H and SAS4) provide a general procedure for cross-section preparation, fuel depletion/decay calculation and general onedimensional or multi-dimensional shielding analysis. The module SAS4 used in the analysis presented in this paper, is a three-dimensional Monte Carlo shielding analysis module, which uses an automated biasing procedure specialized for a nuclear fuel transport or storage container. The Spent Fuel Interim Storage Facility in our country is projected to be a parallelepiped concrete monolithic module, consisting of an external reinforced concrete structure with vertical storage cylinders (pits) arranged in a rectangular array. A pit is filled with sealed cylindrical baskets of stainless steel arranged in a stack, and with each basket containing spent fuel bundles in vertical position. The pit is closed with a concrete plug. The cylindrical geometry model is used in the shielding evaluation for a spent fuel storage structure (pit), and only the active parts of the superposed bundles is considered. The dose rates have been calculated in both the axial and radial directions using SAS4.(author)

  7. Estimated routine radiation doses to transportation workers in alternative spent-fuel transportation systems

    International Nuclear Information System (INIS)

    Schneider, K.J.; Smith, R.I.; Daling, P.M.; Ross, W.A.; McNair, G.W.

    1988-01-01

    The federal system for the management of spent fuel and high-level radioactive waste includes the acceptance by the US Department of Energy (DOE) of the spent fuel or waste loaded in casks at the reactor or other waste generators, its transportation to a repository, and its handling and final emplacement in the repository. The DOE plans to implement a transportation system that is safe, secure, efficient, and cost-effective and will meet applicable regulatory safety and security requirements. The DOE commissioned the Pacific Northwest Laboratory (PNL) to develop estimates of the routine radiation doses that would result from the operation of a system postulated using current designs and practices. From that evaluation, PNL identified activities/operations that result in the higher fraction of doses, proposed conceptual alternatives that would effectively reduce such exposures, and evaluated the cost-effectiveness of such alternatives. The study is one of a series used in making overall system design and operational decisions in the development of the DOE's spent-fuel/high-level waste transportation system. This paper contains the highlights from the PNL study of the estimated radiation doses to the transportation workers in a postulated reference transportation system and potential alternatives to that system

  8. Required doses for projection methods in X-ray diagnosis

    International Nuclear Information System (INIS)

    Hagemann, G.

    1992-01-01

    The ideal dose requirement has been stated by Cohen et al. (1981) by a formula basing on parallel beam, maximum quantum yield and Bucky grid effect depending on the signal to noise ratio and object contrast. This was checked by means of contrast detail diagrams measured at the hole phantom, and was additionally compared with measurement results obtained with acrylic glass phantoms. The optimal dose requirement is obtained by the maximum technically possible approach to the ideal requirement level. Examples are given, besides for x-ray equipment with Gd 2 O 2 S screen film systems for grid screen mammography, and new thoracic examination systems for mass screenings. Finally, a few values concerning the dose requirement or the analogous time required for fluorscent screening in angiography and interventional radiology, are stated, as well as for dentistry and paediatric x-ray diagnostics. (orig./HP) [de

  9. Nuclear fuel particles in the environment - characteristics, atmospheric transport and skin doses

    International Nuclear Information System (INIS)

    Poellaenen, R.

    2002-05-01

    In the present thesis, nuclear fuel particles are studied from the perspective of their characteristics, atmospheric transport and possible skin doses. These particles, often referred to as 'hot' particles, can be released into the environment, as has happened in past years, through human activities, incidents and accidents, such as the Chernobyl nuclear power plant accident in 1986. Nuclear fuel particles with a diameter of tens of micrometers, referred to here as large particles, may be hundreds of kilobecquerels in activity and even an individual particle may present a quantifiable health hazard. The detection of individual nuclear fuel particles in the environment, their isolation for subsequent analysis and their characterisation are complicated and require well-designed sampling and tailored analytical methods. In the present study, the need to develop particle analysis methods is highlighted. It is shown that complementary analytical techniques are necessary for proper characterisation of the particles. Methods routinely used for homogeneous samples may produce erroneous results if they are carelessly applied to radioactive particles. Large nuclear fuel particles are transported differently in the atmosphere compared with small particles or gaseous species. Thus, the trajectories of gaseous species are not necessarily appropriate for calculating the areas that may receive large particle fallout. A simplified model and a more advanced model based on the data on real weather conditions were applied in the case of the Chernobyl accident to calculate the transport of the particles of different sizes. The models were appropriate in characterising general transport properties but were not able to properly predict the transport of the particles with an aerodynamic diameter of tens of micrometers, detected at distances of hundreds of kilometres from the source, using only the current knowledge of the source term. Either the effective release height has been higher

  10. Nuclear fuel particles in the environment - characteristics, atmospheric transport and skin doses

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R

    2002-05-01

    In the present thesis, nuclear fuel particles are studied from the perspective of their characteristics, atmospheric transport and possible skin doses. These particles, often referred to as 'hot' particles, can be released into the environment, as has happened in past years, through human activities, incidents and accidents, such as the Chernobyl nuclear power plant accident in 1986. Nuclear fuel particles with a diameter of tens of micrometers, referred to here as large particles, may be hundreds of kilobecquerels in activity and even an individual particle may present a quantifiable health hazard. The detection of individual nuclear fuel particles in the environment, their isolation for subsequent analysis and their characterisation are complicated and require well-designed sampling and tailored analytical methods. In the present study, the need to develop particle analysis methods is highlighted. It is shown that complementary analytical techniques are necessary for proper characterisation of the particles. Methods routinely used for homogeneous samples may produce erroneous results if they are carelessly applied to radioactive particles. Large nuclear fuel particles are transported differently in the atmosphere compared with small particles or gaseous species. Thus, the trajectories of gaseous species are not necessarily appropriate for calculating the areas that may receive large particle fallout. A simplified model and a more advanced model based on the data on real weather conditions were applied in the case of the Chernobyl accident to calculate the transport of the particles of different sizes. The models were appropriate in characterising general transport properties but were not able to properly predict the transport of the particles with an aerodynamic diameter of tens of micrometers, detected at distances of hundreds of kilometres from the source, using only the current knowledge of the source term. Either the effective release height has

  11. Review of radionuclides released from the nuclear fuel cycle and methods of assessing dose to man

    International Nuclear Information System (INIS)

    Bryant, P.M.

    1979-01-01

    There are two broad subject areas associated with releases of radionuclides from nuclear fuel cycle installations to the environment in which there are biological implications. One concerns interpretation of doses to man in terms of their radiological significance; the other concerns estimation of environmental transfer of radionuclides and of associated radiation doses to man. The radiation protection philosophy on which past practice regarding effluent releases of radionuclides to the environment was based is illustrated by drawing upon estimates of the associated radiation doses to man given in the 1977 report of the United Nations Scientific Committee on the Effects of Atomic Radiation. The present emphasis in radiation protection philosophy is illustrated by summarizing a review of environmental models relevant to estimation of radiation doses to population groups with reference to effluent releases of 3 H, 14 C, 85 Kr and 129 I; the author carried out the review as a contribution to a current study by an expert group set up by the Nuclear Energy Agency of OECD. Radionuclides of significance in the future may differ from those currently released to the environment because of possible developments in nuclear fuel cycles and options which may be exercised for disposal of high-level radioactive wastes, already in storage or postulated to be produced in the future. (author)

  12. Major results and lessons learned for performance assessments of spent fuel geological disposal: the SPA project

    International Nuclear Information System (INIS)

    Baudoin, P.; Serres, C.; Certes, C.; Gay, D.

    2001-01-01

    This paper presents a summary of the results obtained in the framework of the SPA (spent fuel disposal performance assessment) project. The project was undertaken by ENRESA, E; GRS, D; IPSN, F; NRG, NL; SCK.CEN, B and VTT, FIN between May 1996 and April 1999. Devoted to the study of spent fuel disposal in various host rock formations (clay, crystalline rocks and salt formation), it notably had the objective to evaluate the long-term performance of different repository systems and to identify the most influential elements. The variety of concepts, sites and scenarios considered in the framework of this project provides a wide range of information from which some general conclusions can be drawn. Focusing on the work done in the case of granite host rock formations, this paper describes the various approaches adopted and states the main sources of differences. It particularly stresses the differences related to the geosphere and biosphere modelling. For the geosphere modelling, ENRESA, GRS and VTT use one dimensional discrete approaches to model the migration of contaminants through the geosphere taking into account for matrix diffusion, whereas IPSN uses a three dimensional continuum approach based on a single porosity model. The comparison of the biosphere conversion factors shows the high influence on the calculated radionuclide dose contributions that can results from biosphere modelling assumptions. It notably points out the differences existing between a simplified ''water drinking'' approach as implemented by VTT and a more classical one in which a wider range of exposure pathways are taken into account. (orig.)

  13. Food-chain and dose model, CALDOS, for assessing Canada's Nuclear Fuel Waste Management concept

    International Nuclear Information System (INIS)

    Zach, R.; Sheppard, S.C.

    1991-01-01

    The food-chain and dose model, CALculation of DOSe (CALDOS), was developed for assessing Canada's concept for nuclear fuel waste disposal in a vault deep in crystalline rock of the Canadian Shield. The model is very general and based on the Shield as a whole. The critical group is totally self-sufficient and represented by ICRP (1975) Reference Man for dose prediction. CALDOS assumes steady-state conditions and deals with variation and uncertainty through Monte Carlo simulation techniques. Ingrowth of some radioactive daughters is considered during food-chain transfer. A limit is set on root uptake to avoid unrealistic plant concentrations. Integrated ingestion and inhalation rates of man are calculated in a unique way, based on energy needs. Soil ingestion by man and external exposure from building material are unique pathways considered. Tritium, 129 I, and 222 Rn are treated through special models, and 14 C and 129 I involve unique geosphere dose limits. All transfer coefficients are lognormally distributed, and the plant/soil concentration ratio is correlated with the soil partition coefficient. Animals' ingestion rates are normally distributed and correlated with each other. Comprehensive sets of internal and external dose conversion factors were calculated for CALDOS. Sample calculations show that dose distributions tend to be strongly right-skewed. Many features of CALDOS are relevant for environmental assessment in general

  14. Potential effects of climatic change on radiological doses from disposal of Canadian nuclear fuel waste

    International Nuclear Information System (INIS)

    Amiro, B.D.

    1997-01-01

    The environmental assessment of deep geologic disposal of Canadian nuclear fuel waste considers many processes that could affect radionuclide transport to humans over thousands of years. Climatic change is an important feature that will occur over these long times. Glaciation will likely occur within the next 100,000 years over much of Canada, and its impact on radiological doses has been assessed previously. In the present study, we investigate the potential effect of short- term climatic change, usually associated with global warming caused by increases in atmospheric trace gases. We study the main biosphere transport pathways causing a radiological dose to humans from 129 I, which is the most important radionuclide in disposal of Canadian used nuclear fuel. Irrigation of a garden with contaminated well water is the main pathway and it can be affected by changes in temperature and precipitation. A cold, wet climate decreases the need for irrigation, and this decreases the radiological dose. A drastic climatic change, such as an increase in temperature from 10 to 20 degrees C and decrease in precipitation from 0.3 to 0.2 m during the growing season, is estimated to increase the dose by a factor of four. This is a relatively small change compared to the range of doses that arise from the variability and uncertainty in many of the parameters used in the environmental assessment models. Therefore, it is likely that the results of probabilistic dose assessment models can include the consequences of short-term climatic change. 39 refs., 3 figs

  15. Numerical Studies on Controlling Gaseous Fuel Combustion by Managing the Combustion Process of Diesel Pilot Dose in a Dual-Fuel Engine

    Directory of Open Access Journals (Sweden)

    Mikulski Maciej

    2015-06-01

    Full Text Available Protection of the environment and counteracting global warming require finding alternative sources of energy. One of the methods of generating energy from environmentally friendly sources is increasing the share of gaseous fuels in the total energy balance. The use of these fuels in compression-ignition (CI engines is difficult due to their relatively high autoignition temperature. One solution for using these fuels in CI engines is operating in a dualfuel mode, where the air and gas mixture is ignited with a liquid fuel dose. In this method, a series of relatively complex chemical processes occur in the engine's combustion chamber, related to the combustion of individual fuel fractions that interact with one another. Analysis of combustion of specific fuels in this type of fuel injection to the engine is difficult due to the fact that combustion of both fuel fractions takes place simultaneously. Simulation experiments can be used to analyse the impact of diesel fuel combustion on gaseous fuel combustion. In this paper, we discuss the results of simulation tests of combustion, based on the proprietary multiphase model of a dual-fuel engine. The results obtained from the simulation allow for analysis of the combustion process of individual fuels separately, which expands the knowledge obtained from experimental tests on the engine.

  16. Fuel supply strategy for power projects in Asia

    International Nuclear Information System (INIS)

    Boyd, D.

    1997-01-01

    Electricity generation is forecast to grow in Asia over the next twenty years. This will bring with it a significant growth in demand for energy resources. Major Asian utility coal procurement strategy to date has aimed at securing a balance between two conflicting needs - security of supply and low cost of supply. Despite the oversupply of coal in the eighties, the two countries experiencing significant growth in demand, Japan and Korea, planned their fuel strategies to ensure security of supply. The supply shortage over the past two years has justified their forward thinking in keeping long term coal contracts in place. Changes are already happening. Recent developments are seeing a move towards the positive attributes of low ash/low sulphur coals appropriately focusing suppliers thoughts on the need in the future to develop and concentrate on environmentally attractive coals to achieve greater returns. Indeed, generally there is a trend in North Asia to demand coal with lower sulphur, and lower ash. The aim now should be to persuade consumers to amend the narrow benchmark view in favour of one that considers the value in use of a particular coal to the customer. The method of pricing thermal coals is also likely to be influenced by the liberalization of the electricity market itself. The introduction of competition amongst generators and the appearance and growth of the Independent Power Producers could well see further pressure on the current simplistic benchmark system. Those financing such projects will seek to reduce risk by every means possible and, in particular, will want to be confident of the long term security of coal supplies

  17. Intercomparison exercise on internal dose assessment. Final report of a joint IAEA-IDEAS project

    International Nuclear Information System (INIS)

    2007-09-01

    There have been several intercomparison exercises organized already at national and international levels for the assessment of occupational exposure due to intakes of radionuclides. These intercomparison exercises revealed significant differences in approaches, methods and assumptions, and consequently in the results. Because of the relevance of the issue for internal dosimetrists, the IAEA organized a new intercomparison exercise in cooperation with the IDEAS project General Guidelines for the Evaluation of Incorporation Monitoring Data, launched under the 5th EU Framework Programme (EU Contract No. FIKR-CT2001-00160). This new intercomparison exercise focused especially on the effect of the guidelines for harmonization of internal dosimetry. It also considered the following aspects: - to provide possibilities for the participating laboratories to check the quality of their internal dose assessment methods in applying the recent ICRP recommendations (e.g. for the new respiratory tract model); - to compare different approaches in interpretation of internal contamination monitoring data; - to quantify the differences in internal dose assessments based on the new guidelines or on other procedures, respectively; - to provide some figures for the influence of the input parameters on the monitoring results; and - to provide a broad forum for information exchange. Several cases have been selected for this exercise with the aim of covering a wide range of practices in the nuclear fuel cycle and in medical applications. The cases were: 1. Acute intake of HTO; 2. Acute inhalation of fission products 137 Cs and 90 Sr; 3. Intake of 60 Co; 4. Repeated intakes of 131 I; 5. Intake of enriched uranium; 6. Single intake of plutonium radionuclides and 241 Am. An Internet based approach had been used for the presentation of the cases, collection of responses and potential discussion of the results. Solutions to these cases were reported by 80 participants worldwide. This report

  18. Correlation between scatter radiation dose at the height of the operators eye and dose to patient for different angiographies projections

    Energy Technology Data Exchange (ETDEWEB)

    Leyton, F.; Nogueira, M. S.; Da Silva, T. A. [Centro de Desenvolvimento da Tecnologia Nuclear / CNEN, Post-graduation in Sciences and Technology of Radiations, Minerals and Materials, Pte. Antonio Carlos No. 6627, Belo Horizonte 31270-901, Minas Gerais (Brazil); Gubolino, L.; Pivetta, M. R. [Hospital dos Fornecedores de Cana de Piracicaba, Av. Barao de Valenca 616, 13405-233 Piracicaba (Brazil); Ubeda, C., E-mail: leyton.fernando@gmail.com [Tarapaca University, Health Sciences Faculty, Radiological Sciences Center, Av. Gral. Velasquez 1775, 1000007 Arica, Arica and Parinacota (Chile)

    2015-10-15

    Cases of radiation induced cataract among cardiology professionals have been reported in studies. In view of evidence of radiation injuries, the ICRP recommends limiting the radiation dose to the lens to 20 mSv per year for occupational exposure. The aim of this works was to report scattered radiation doses at the height of the operators eye in an interventional cardiology facility from procedures performed without use of radiation protection devices, correlated with different angiographic projections and operational modes. Measurements were made in a cardiac laboratory with an angiography X-ray system GE equipped with flat-panel detector. PMMA plates of 30 x 30 x 5 cm were used to simulate a patient with a thickness of 20 cm. Two fluoroscopy modes (low and normal, 15 frame/s), cine mode 15 frame/s. Four angiographic projections anterior posterior (Ap), lateral (Lat), left anterior oblique caudal (spider) and left anterior oblique cranial (Lao-45/cra-30) and a cardiac protocol for patient between 70 to 90 kg was used. Measurements of phantom entrance doses rate and scatter doses rate were performed with two Unfors Xi plus. The detector measuring scatter radiation was positioned at the usual distance of the cardiologists eyes during working conditions (1 m from the isocenter and 1.7 m from the floor). There is a good linear correlation between the kerma-area product and scatter dose at the lens. An experimental correlation factor of 2.3; 12.0; 12.2 and 17.6 μSv/Gy cm{sup 2} were found for the Ap, Lao/cra, spider and Lat projections, respectively. The entrance dose of PMMA for fluoroscopy low, medium and cine was 13, 39 and 282 mGy/min, respectively to Ap. (Author)

  19. Correlation between scatter radiation dose at the height of the operators eye and dose to patient for different angiographies projections

    International Nuclear Information System (INIS)

    Leyton, F.; Nogueira, M. S.; Da Silva, T. A.; Gubolino, L.; Pivetta, M. R.; Ubeda, C.

    2015-10-01

    Cases of radiation induced cataract among cardiology professionals have been reported in studies. In view of evidence of radiation injuries, the ICRP recommends limiting the radiation dose to the lens to 20 mSv per year for occupational exposure. The aim of this works was to report scattered radiation doses at the height of the operators eye in an interventional cardiology facility from procedures performed without use of radiation protection devices, correlated with different angiographic projections and operational modes. Measurements were made in a cardiac laboratory with an angiography X-ray system GE equipped with flat-panel detector. PMMA plates of 30 x 30 x 5 cm were used to simulate a patient with a thickness of 20 cm. Two fluoroscopy modes (low and normal, 15 frame/s), cine mode 15 frame/s. Four angiographic projections anterior posterior (Ap), lateral (Lat), left anterior oblique caudal (spider) and left anterior oblique cranial (Lao-45/cra-30) and a cardiac protocol for patient between 70 to 90 kg was used. Measurements of phantom entrance doses rate and scatter doses rate were performed with two Unfors Xi plus. The detector measuring scatter radiation was positioned at the usual distance of the cardiologists eyes during working conditions (1 m from the isocenter and 1.7 m from the floor). There is a good linear correlation between the kerma-area product and scatter dose at the lens. An experimental correlation factor of 2.3; 12.0; 12.2 and 17.6 μSv/Gy cm 2 were found for the Ap, Lao/cra, spider and Lat projections, respectively. The entrance dose of PMMA for fluoroscopy low, medium and cine was 13, 39 and 282 mGy/min, respectively to Ap. (Author)

  20. Connecticut nutmeg fuel cell bus project : first analysis report.

    Science.gov (United States)

    2012-07-01

    This report summarizes the experience and early results from a fuel cell bus demonstration funded by the Federal Transit Administra-tion (FTA) under the National Fuel Cell Bus Program (NFCBP). A team led by the Northeast Advanced Vehicle Consortium a...

  1. EBR-II spent fuel treatment demonstration project

    International Nuclear Information System (INIS)

    Benedict, R.W.; Henslee, S.P.

    1997-01-01

    For approximately 10 years, Argonne National Laboratory was developed a fast reactor fuel cycle based on dry processing. When the US fast reactor program was canceled in 1994, the fuel processing technology, called the electrometallurgical technique, was adapted for treating unstable spent nuclear fuel for disposal. While this technique, which involves electrorefining fuel in a molten salt bath, is being developed for several different fuel categories, its initial application is for sodium-bonded metallic spent fuel. In June 1996, the Department of Energy (DOE) approved a radiation demonstration program in which 100 spent driver assemblies and 25 spent blanket assemblies from the Experimental Breeder Reactor-II (EBR-II) will be treated over a three-year period. This demonstrated will provide data that address issues in the National Research Council's evaluation of the technology. The planned operations will neutralize the reactive component (elemental sodium) in the fuel and produce a low enriched uranium product, a ceramic waste and a metal waste. The fission products and transuranium elements, which accumulate in the electrorefining salt, will be stabilized in the glass-bonded ceramic waste form. The stainless steel cladding hulls, noble metal fission products, and insoluble residues from the process will be stabilized in a stainless steel/zirconium alloy. Upon completion of a successful demonstration and additional environmental evaluation, the current plans are to process the remainder of the DOE sodium bonded fuel

  2. Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC

  3. Oxide fuel fabrication technology development of the FaCT project (1). Overall review of fuel technology development of the FaCT project

    International Nuclear Information System (INIS)

    Abe, Tomoyuki; Namekawa, Takashi; Tanaka, Kenya

    2011-01-01

    The FaCT project is in progress in Japan for the commercialization of fast reactor cycle system. The development goal of the fuel in the FaCT project is a low-decontaminated TRU homo-recycling in a closed cycle and extension in average discharge burn-up to 150 GWd/t. Research and development on innovative technologies concerning the short process, remote maintenance and cooling system of automatic fuel production equipments, long life cladding material and control of oxygen potential have been conducted in phase I of the FaCT project. As the result of various test including 600 g batch MOX tests, it is concluded that the short process is available to fuel pellet fabrication of the FaCT project. Although cold mock-up tests on test model of some typical process equipments suggest possibilities of remote maintenance of automatic fuel fabrication equipment, it is concluded that it still needs further efforts to judge the operability of the completely remote fabrication for low-decontaminated TRU fuel. A cold mock-up test on fuel pin assembling equipment show that influence of decay heat of MA can be managed by cooling system. Irradiation tests in BOR-60 indicate that 9Cr-ODS possess the satisfactory in-reactor performance as the long life cladding material if homogeneity of alloy element is adequately controlled. Modification of cladding tube fabrication process to ensure homogeneity and further development of measures to control oxygen potential inside the fuel pin are necessary to reach the burn-up target of the FaCT project. (author)

  4. Decay heat and gamma dose-rate prediction capability in spent LWR fuel

    International Nuclear Information System (INIS)

    Neely, G.J.; Schmittroth, F.

    1982-08-01

    The ORIGEN2 code was established as a valid means to predict decay heat from LWR spent fuel assemblies for decay times up to 10,000 year. Calculational uncertainties ranged from 8.6% to a maximum of 16% at 2.5 years and 300 years cooling time, respectively. The calculational uncertainties at 2.5 years cooling time are supported by experiment. Major sources of uncertainty at the 2.5 year cooling time were identifed as irradiation history (5.7%) and nuclear data together with calculational methods (6.3%). The QAD shielding code was established as a valid means to predict interior and exterior gamma dose rates of spent LWR fuel assemblies. A calculational/measurement comparison was done on two assemblies with different irradiation histories and supports a 35% calculational uncertainty at the 1.8 and 3.0 year decay times studied. Uncertainties at longer times are expected to increase, but not significantly, due to an increased contribution from the actinides whose inventories are assigned a higher uncertainty. The uncertainty in decay heat rises to a maximum of 16% due to actinide uncertainties. A previous study was made of the neutron emission rate from a typical Turkey Point Unit 3, Region 4 spent fuel assembly at 5 years decay time. A conservative estimate of the neutron dose rate at the assembly surface was less than 0.5 rem/hr

  5. Anticipated Radiological Dose to Worker for Plutonium Stabilization and Handling at PFP - Project W-460

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

    This report provides estimates of the expected whole body and extremity radiological dose, expressed as dose equivalent (DE), to workers conducting planned plutonium (Pu) stabilization processes at the Hanford Site Plutonium Finishing Plant (PFP). The report is based on a time and motion dose study commissioned for Project W-460, Plutonium Stabilization and Handling, to provide personnel exposure estimates for construction work in the PFP storage vault area plus operation of stabilization and packaging equipment at PFP

  6. Anticipated Radiological Dose to Worker for Plutonium Stabilization and Handling at PFP - Project W-460

    CERN Document Server

    Weiss, E V

    2000-01-01

    This report provides estimates of the expected whole body and extremity radiological dose, expressed as dose equivalent (DE), to workers conducting planned plutonium (Pu) stabilization processes at the Hanford Site Plutonium Finishing Plant (PFP). The report is based on a time and motion dose study commissioned for Project W-460, Plutonium Stabilization and Handling, to provide personnel exposure estimates for construction work in the PFP storage vault area plus operation of stabilization and packaging equipment at PFP.

  7. Dose-projection considerations for emergency conditions at nuclear power plants

    International Nuclear Information System (INIS)

    Stoetzel, G.A.; Ramsdell, J.V.; Poeton, R.W.; Powell, D.C.; Desrosiers, A.E.

    1983-05-01

    The purpose of this report is to review the problems and issues associated with making environmental radiation-dose projections during emergencies at nuclear power plants. The review is divided into three areas: source-term development, characterization of atmospheric dispersion and selection of appropriate dispersion models, and development of dosimetry calculations for determining thyroid dose and whole-body dose for ground-level and elevated releases. A discussion of uncertainties associated with these areas is also provided

  8. Dose-projection considerations for emergency conditions at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Stoetzel, G.A.; Ramsdell, J.V.; Poeton, R.W.; Powell, D.C.; Desrosiers, A.E.

    1983-05-01

    The purpose of this report is to review the problems and issues associated with making environmental radiation-dose projections during emergencies at nuclear power plants. The review is divided into three areas: source-term development, characterization of atmospheric dispersion and selection of appropriate dispersion models, and development of dosimetry calculations for determining thyroid dose and whole-body dose for ground-level and elevated releases. A discussion of uncertainties associated with these areas is also provided.

  9. Effect of Localizer Radiography Projection on Organ Dose at Chest CT with Automatic Tube Current Modulation.

    Science.gov (United States)

    Saltybaeva, Natalia; Krauss, Andreas; Alkadhi, Hatem

    2017-03-01

    Purpose To calculate the effect of localizer radiography projections to the total radiation dose, including both the dose from localizer radiography and that from subsequent chest computed tomography (CT) with tube current modulation (TCM). Materials and Methods An anthropomorphic phantom was scanned with 192-section CT without and with differently sized breast attachments. Chest CT with TCM was performed after one localizer radiographic examination with anteroposterior (AP) or posteroanterior (PA) projections. Dose distributions were obtained by means of Monte Carlo simulations based on acquired CT data. For Monte Carlo simulations of localizer radiography, the tube position was fixed at 0° and 180°; for chest CT, a spiral trajectory with TCM was used. The effect of tube start angles on dose distribution was investigated with Monte Carlo simulations by using TCM curves with fixed start angles (0°, 90°, and 180°). Total doses for lungs, heart, and breast were calculated as the sum of the dose from localizer radiography and CT. Image noise was defined as the standard deviation of attenuation measured in 14 circular regions of interest. The Wilcoxon signed rank test, paired t test, and Friedman analysis of variance were conducted to evaluate differences in noise, TCM curves, and organ doses, respectively. Results Organ doses from localizer radiography were lower when using a PA instead of an AP projection (P = .005). The use of a PA projection resulted in higher TCM values for chest CT (P chest CT. © RSNA, 2016 Online supplemental material is available for this article.

  10. Preliminary design review report for K Basin Dose Reduction Project

    International Nuclear Information System (INIS)

    Blackburn, L.D.

    1996-01-01

    The strategy for reducing radiation dose, originating from radionuclides absorbed in the K East Basin concrete, is to raise the pool water level to provide additional shielding. This report documents a preliminary design review conducted to ensure that design approaches for cleaning/coating basin walls and modifying other basin components were appropriate. The conclusion of this review was that design documents presently conclusion of this review was that design documents presently completed or in process of modification are and acceptable basis for proceeding to complete the design

  11. Quality Assurance Program Plan for Project W-379: Spent Nuclear Fuels Canister Storage Building Projec

    International Nuclear Information System (INIS)

    Duncan, D.W.

    1995-01-01

    This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner

  12. OECD-IAEA Paks Fuel Project. Detailed Description of the Results of Calculations

    International Nuclear Information System (INIS)

    2010-05-01

    On 10 April 2003 severe damage of fuel assemblies took place during an incident at Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a special tank below the water level of the spent fuel storage pool in order to remove crud buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods were being cooled by circulation of storage pool water. The first sign of fuel failure was the detection of some fission gases released from the cleaning tank during that evening. The cleaning tank cover locks were released after midnight and this operation was followed by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel assemblies were severely damaged. The first evaluation of the event showed that the severe fuel damage happened due to inadequate coolant circulation within the cleaning tank. The damaged fuel assemblies will be removed from the cleaning tank in 2005 and will be stored in special canisters in the spent fuel storage pool of the Paks NPP. Following several discussions between expert from different countries and international organisations the OECD-IAEA Paks Fuel Project was proposed. The project is envisaged in two phases. - Phase 1 is to cover organization of visual inspection of material, preparation of database, performance of analyses and preparatory work for fuel examination. - Phase 2 is to cover the fuel transport and the hot cell examination

  13. DE-NE0000735 - FINAL REPORT ON THORIUM FUEL CYCLE NEUP PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    Krahn, Steven [Vanderbilt Univ., Nashville, TN (United States); Ault, Timothy [Vanderbilt Univ., Nashville, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-30

    The report is broken into six chapters, including this executive summary chapter. Following an introduction, this report discusses each of the project’s three major components (Fuel Cycle Data Package (FCDP) Development, Thorium Fuel Cycle Literature Analysis and Database Development, and the Thorium Fuel Cycle Technical Track and Proceedings). A final chapter is devoted to summarization. Various outcomes, publications, etc. originating from this project can be found in the Appendices at the end of the document.

  14. Comparison of neutron dose measured by Albedo TLD and etched tracks detector at PNC plutonium fuel facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Momose, T.; Shinohara, K.; Ishiguro, H.

    1996-01-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) has fabricated Plutonium and Uranium Mixed OXide (MOX) fuel for FBR MONJU at Tokai works. In this site, PNC/Panasonic albedo TLDs/1/ are used for personnel neutron monitoring. And a part of workers wore Etched Tracks Detector (ETD) combined with TLD in order to check the accuracy of the neutron dose estimated by albedo TLD. In this paper, the neutron dose measured by TLD and ETD in the routine monitoring is compared at PNC plutonium fuel facilities. (author)

  15. Spent fuel waste disposal: analyses of model uncertainty in the MICADO project

    International Nuclear Information System (INIS)

    Grambow, B.; Ferry, C.; Casas, I.; Bruno, J.; Quinones, J.; Johnson, L.

    2010-01-01

    The objective was to find out whether international research has now provided sufficiently reliable models to assess the corrosion behavior of spent fuel in groundwater and by this to contribute to answering the question whether the highly radioactive used fuel from nuclear reactors can be disposed of safely in a geological repository. Principal project results are described in the paper

  16. Fuel Rod Consolidation Project: Phase 2, Final report: Volume 2, Appendices

    International Nuclear Information System (INIS)

    1987-01-01

    This document, Volume 2, provides the appendices to Volume 1 of the Fuel Rod Consolidation Project. It provides information on the following: References; Trade-off Studies; Instrument List; RAM Data; Fabrication Specifications; Software Specifications; and Design Requirements

  17. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    International Nuclear Information System (INIS)

    MITCHELL, R.M.

    2000-01-01

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey

  18. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    2000-10-12

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  19. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    2000-09-28

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  20. Dose prediction for plants and animals - nuclear fuel waste management perspective

    International Nuclear Information System (INIS)

    Zach, R.; Amiro, B.

    1997-01-01

    We have developed a comprehensive, practical ecological radiation assessment methodology and applied it in the environmental impact statement (EIS) for evaluating the safety of Canada's nuclear fuel waste disposal concept. The methodology has four screening steps, and we focus here on the last two concerned with dose estimation for . plants and animals. We present ten classes of issues that were compiled from comments regarding our methodology from EIS review participants. Furthermore, we identify future needs and developments for improving our methodology. The issues raised by EIS participants, and the future needs and developments indicated by us are also of general importance in guiding future work. (author)

  1. Fuel Cell-Powered Lift Truck Fleet Deployment Projects Final Technical Report May 2014

    Energy Technology Data Exchange (ETDEWEB)

    Klingler, James J [GENCO Infrastructure Solutions, Inc.

    2014-05-06

    The overall objectives of this project were to evaluate the performance, operability and safety of fork lift trucks powered by fuel cells in large distribution centers. This was accomplished by replacing the batteries in over 350 lift trucks with fuel cells at five distribution centers operated by GENCO. The annual cost savings of lift trucks powered by fuel cell power units was between $2,400 and $5,300 per truck compared to battery powered lift trucks, excluding DOE contributions. The greatest savings were in fueling labor costs where a fuel cell powered lift truck could be fueled in a few minutes per day compared to over an hour for battery powered lift trucks which required removal and replacement of batteries. Lift truck operators where generally very satisfied with the performance of the fuel cell power units, primarily because there was no reduction in power over the duration of a shift as experienced with battery powered lift trucks. The operators also appreciated the fast and easy fueling compared to the effort and potential risk of injury associated with switching heavy batteries in and out of lift trucks. There were no safety issues with the fueling or operation of the fuel cells. Although maintenance costs for the fuel cells were higher than for batteries, these costs are expected to decrease significantly in the next generation of fuel cells, making them even more cost effective.

  2. Integrated task plans for the Hanford Environmental Dose Reconstruction Project, June 1992 through May 1994

    International Nuclear Information System (INIS)

    Shipler, D.B.

    1993-09-01

    The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses from Hanford Site operations since 1944 to representative individuals. The primary objective of work to be performed through May 1994 is to determine the project's appropriate scope: space, time, radionuclides, pathways and representative individuals; determine the project's appropriate level of accuracy/level of uncertainty in dose estimates; complete model and data development; and estimate doses for the Hanford Thyroid Disease Study and representative individuals. A major objective of the HEDR Project is to estimate doses to the thyroid of individuals who were exposed to iodine-131. A principal pathway for many of these individuals was milk from cows that ate vegetation contaminated by iodine-131 released into the air from Hanford facilities. The plan for June 1992 through May 1994 has been prepared based on activities and budgets approved by the Technical Steering Panel (TSP) at its meetings on January 7--9, 1993 and February 25--26, 1993. The activities can be divided into three broad categories: (1) computer code and data development activities, (2) calculation of doses, and (3) technical and communication support to the TSP and the TSP Native American Working Group (NAWG). The following activities will be conducted to accomplish project objectives through May 1994

  3. Columbia River pathway report: phase I of the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    1991-07-01

    This report summarizes the river-pathway portion of the first phase of the Hanford Environmental Dose Reconstruction (HEDR) Project. The HEDR Project is estimating radiation doses that could have been received by the public from the Department of Energy's Hanford Site, in southeastern Washington State. Phase 1 of the river-pathway dose reconstruction effort sought to determine whether dose estimates could be calculated for populations in the area from above the Hanford Site at Priest Rapids Dam to below the site at McNary Dam from January 1964 to December 1966. Of the potential sources of radionuclides from the river, fish consumption was the most important. Doses from drinking water were lower at Pasco than at Richland and lower at Kennewick than at Pasco. The median values of preliminary dose estimates calculated by HEDR are similar to independent, previously published estimates of average doses to Richland residents. Later phases of the HEDR Project will address dose estimates for periods other than 1964--1966 and for populations downstream of McNary Dam. 17 refs., 19 figs., 1 tab.

  4. A simple method to back-project isocenter dose of radiotherapy treatments using EPID transit dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Silveira, T.B.; Cerbaro, B.Q.; Rosa, L.A.R. da, E-mail: thiago.fisimed@gmail.com, E-mail: tbsilveira@inca.gov.br [Instituto de Radioproteção e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro - RJ (Brazil)

    2017-07-01

    The aim of this work was to implement a simple algorithm to evaluate isocenter dose in a phantom using the back-projected transmitted dose acquired using an Electronic Portal Imaging Device (EPID) available in a Varian Trilogy accelerator with two nominal 6 and 10 MV photon beams. This algorithm was developed in MATLAB language, to calibrate EPID measured dose in absolute dose, using a deconvolution process, and to incorporate all scattering and attenuation contributions due to photon interactions with phantom. Modeling process was simplified by using empirical curve adjustments to describe the contribution of scattering and attenuation effects. The implemented algorithm and method were validated employing 19 patient treatment plans with 104 clinical irradiation fields projected on the phantom used. Results for EPID absolute dose calibration by deconvolution have showed percent deviations lower than 1%. Final method validation presented average percent deviations between isocenter doses calculated by back-projection and isocenter doses determined with ionization chamber of 1,86% (SD of 1,00%) and -0,94% (SD of 0,61%) for 6 and 10 MV, respectively. Normalized field by field analysis showed deviations smaller than 2% for 89% of all data for 6 MV beams and 94% for 10 MV beams. It was concluded that the proposed algorithm possesses sufficient accuracy to be used for in vivo dosimetry, being sensitive to detect dose delivery errors bigger than 3-4% for conformal and intensity modulated radiation therapy techniques. (author)

  5. Solubility of hot fuel particles from Chernobyl--influencing parameters for individual radiation dose calculations.

    Science.gov (United States)

    Garger, Evgenii K; Meisenberg, Oliver; Odintsov, Oleksiy; Shynkarenko, Viktor; Tschiersch, Jochen

    2013-10-15

    Nuclear fuel particles of Chernobyl origin are carriers of increased radioactivity (hot particles) and are still present in the atmosphere of the Chernobyl exclusion zone. Workers in the zone may inhale these particles, which makes assessment necessary. The residence time in the lungs and the transfer in the blood of the inhaled radionuclides are crucial for inhalation dose assessment. Therefore, the dissolution of several kinds of nuclear fuel particles from air filters sampled in the Chernobyl exclusion zone was studied. For this purpose filter fragments with hot particles were submersed in simulated lung fluids (SLFs). The activities of the radionuclides (137)Cs, (90)Sr, (239+240)Pu and (241)Am were measured in the SLF and in the residuum of the fragments by radiometric methods after chemical treatment. Soluble fractions as well as dissolution rates of the nuclides were determined. The influence of the genesis of the hot particles, represented by the (137)Cs/(239+240)Pu ratio, on the availability of (137)Cs was demonstrated, whereas the dissolution of (90)Sr, (239+240)Pu and (241)Am proved to be independent of genesis. No difference in the dissolution of (137)Cs and (239+240)Pu was observed for the two applied types of SLF. Increased solubility was found for smaller hot particles. A two-component exponential model was used to describe the dissolution of the nuclides as a function of time. The results were applied for determining individual inhalation dose coefficients for the workers at the Chernobyl construction site. Greater dose coefficients for the respiratory tract and smaller coefficients for the other organs were calculated (compared to ICRP default values). The effective doses were in general lower for the considered radionuclides, for (241)Am even by one order of magnitude. © 2013 Elsevier B.V. All rights reserved.

  6. Progress and status of the international project on innovative nuclear reactors and fuel cycles (INPRO) - 5182

    International Nuclear Information System (INIS)

    Ponomarev, A.; Fesenko, G.; Grigoriev, F.G.; Korinny, A.; Phillips, J.R.; Rho, K.

    2015-01-01

    The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000 through IAEA General Conference resolution. INPRO cooperates with Member States to ensure that sustainable nuclear energy is available to help meet the energy needs of the 21. century. INPRO membership has grown to 41 members and 16 observers. The paper presents the current prospectus of the INPRO programme and details the most recent achievements in the following 7 projects: 1) the GAINS project (Global Architecture of Innovative Nuclear Energy Systems with thermal and fast reactors and a closed nuclear fuel cycle); 2) the SYNERGIES project applies and amends the analytical framework developed in GAINS project to examine more specifically the various forms of regional collaboration among nuclear energy suppliers and users; 3) the KIND project (Key Indicators for Innovative Nuclear Energy Systems) has the objective of developing guidance on the evaluation on innovative nuclear technologies; 4) the ROADMAPS project addresses several possible stages toward nuclear energy sustainability; 5) the RISC project aims at demonstrating that the evolution of safety requirements and technical innovations provide continual progress towards the avoidance of evacuation measures outside NPP sites in case of severe accidents; 6) the FANES project has the objective of carrying out feasibility analyses of advanced and innovative fuels for different reactor systems; and 7) the WIRAF project aims at identifying problematic waste from innovative reactor designs and corresponding nuclear fuel cycles

  7. Measurements of eye lens doses in interventional radiology and cardiology: Final results of the ORAMED project

    International Nuclear Information System (INIS)

    Vanhavere, F.; Carinou, E.; Domienik, J.; Donadille, L.; Ginjaume, M.; Gualdrini, G.; Koukorava, C.; Krim, S.; Nikodemova, D.; Ruiz-Lopez, N.; Sans-Merce, M.; Struelens, L.

    2011-01-01

    Within the ORAMED project (Optimization of Radiation Protection of Medical Staff) a coordinated measurement program for occupationally exposed medical staff was performed in different hospitals in Europe ( (www.oramed-fp7.eu)). The main objective was to obtain a set of standardized data on extremity and eye lens doses for staff involved in interventional radiology and cardiology and to optimize radiation protection. Special attention was given to the measurement of the doses to the eye lenses. In this paper an overview will be given of the measured eye lens doses and the main influence factors for these doses. The measured eye lens doses are extrapolated to annual doses. The extrapolations showed that monitoring of the eye lens should be performed on routine basis.

  8. Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

    International Nuclear Information System (INIS)

    Matijevic, M.; Grgic, D.; Jecmenica, R.

    2016-01-01

    This paper presents comparison of the Krsko Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectra with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters (SB cards in SDEF and WWINP file) for MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FW-CADIS methodology. Denovo implements a structured, Cartesian-grid SN solver based on the Koch-Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was first

  9. Internal dose evaluation from actinide intakes during nuclear power reactor spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pawar, S.K.; Kumar, Ranjeet; Gamre, Rupali; Purohit, R.G.

    2011-01-01

    Full text: Indian PHWR reactors are using natural uranium as fuel. After use they are discharged from the core and send for fuel reprocessing to extract the unused uranium and plutonium. Plutonium and other actinides are formed by activation of 238 U with neutrons and subsequent decay. During reprocessing of the spent fuel, major long lived actinides (Pu, Am and U) may become radiological safety hazard. Actinides intakes are more probable during declading and chopping of spent fuel. During routine plant operation in reprocessing, exposure to Pu is a major concern along with Am and U in working environment due to its higher radiological hazard and occupational workers are likely to get exposed to plutonium, Americium and Uranium mostly through inhalation. Internally deposited Pu-isotopes, Am-isotope and U-isotopes are estimated using techniques such as lung counting (in-vivo) and urine and faecal bioassay (in-vitro). Evaluation of internal dose of actinides is dependent upon urinary excreted activity. To estimate the internally deposited Pu, U and Am at an intake level of about one ALI (ICRP-78, 1997) of occupational workers, urine bioassay is the preferred technique due to high detection sensitivity, ease of sample handling and economical method. A small and measurable fraction of internally deposited Pu, Am and U are excreted through urine whose content is dependent on time of inhalation, quantity and type of chemical form of inhaled material (S and M class). A standardized radiochemical analysis method for separation and estimation of Pu, Am and U is used to evaluate the urinary excreted activity and internal dose. Several measurements techniques are employed for the estimation of plutonium, Americium and Uranium for example, Alpha Spectrometry, Gamma Spectrometry, Neutron Activation Analysis, Mass Spectrometry and Fission Track Analysis. The radiochemical separation followed by alpha counting and/or spectrometry is chosen due to its ease of handling and

  10. Final Technical Report for the MIT Annular Fuel Research Project

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2008-01-01

    MIT-NFC-PR-082 (January 2006) Abstract This summary provides an overview of the results of the U.S. DOE funded NERI (Nuclear Research Energy Initiative) program on development of the internally and externally cooled annular fuel for high power density PWRs. This new fuel was proposed by MIT to allow a substantial increase in power density (on the order of 30% or higher) while maintaining or improving safety margins. A comprehensive study was performed by a team consisting of MIT (lead organization), Westinghouse Electric Corporation, Gamma Engineering Corporation, Framatome ANP(formerly Duke Engineering) and Atomic Energy of Canada Limited

  11. Numerical analyses of an ex-core fuel incident: Results of the OECD-IAEA Paks Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z., E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Aszodi, A. [BME NTI Budapest (Hungary); Barnak, M. [IVS, Trnava (Slovakia); Boros, I. [BME NTI Budapest (Hungary); Fogel, M. [VUJE, Trnava (Slovakia); Guillard, V. [IRSN, Cadarache (France); Gyori, Cs. [ITU, EU, Karlsruhe (Germany); Hegyi, G. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Horvath, G.L. [VEIKI, Budapest (Hungary); Nagy, I. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Junninen, P. [VTT, Espoo (Finland); Kobzar, V. [KI, Moscow (Russian Federation); Legradi, G. [BME NTI Budapest (Hungary); Molnar, A. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Pietarinen, K. [VTT, Espoo (Finland); Perneczky, L. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Makihara, Y. [ATMEA, Paris (France); Matejovic, P. [IVS, Trnava (Slovakia); Perez-Fero, E.; Slonszki, E. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary)

    2010-03-15

    The OECD-IAEA Paks Fuel Project was developed to support the understanding of fuel behaviour in accident conditions on the basis of analyses of the Paks-2 incident. Numerical simulation of the most relevant aspects of the event and comparison of the calculation results with the available data from the incident was carried out between 2006 and 2007. A database was compiled to provide input for the code calculations. The activities covered the following three areas: (a) Thermal hydraulic calculations described the cooling conditions possibly established during the incident. (b) Simulation of fuel behaviour described the oxidation and degradation mechanisms of the fuel assemblies. (c) The release of fission products from the failed fuel rods was estimated and compared to available measured data. The applied used codes captured the most important events of the Paks-2 incident and the calculated results improved the understanding of the causes and mechanisms of fuel failure. The numerical analyses showed that the by-pass flow leading to insufficient cooling amounted to 75-90% of the inlet flow rate, the maximum temperature in the tank was between 1200 and 1400 deg. C, the degree of zirconium oxidation reached 4-12% and the mass of produced hydrogen was between 3 and 13 kg.

  12. Energizing Engineering Students with Hydrogen Fuel Cell Project

    Science.gov (United States)

    Cannell, Nori; Zavaleta, Dan

    2010-01-01

    At Desert Vista High School, near Phoenix, Arizona, Perkins Innovation Grant funding is being used to fund a program that is helping to prepare students for careers in engineering by giving them hands-on experience in areas like hydrogen generation and fuel cell utilization. As one enters Dan Zavaleta's automotive and engineering classroom and lab…

  13. Spent Nuclear Fuel project systems engineering management plan

    International Nuclear Information System (INIS)

    Womack, J.C.

    1995-01-01

    The purpose of the WHC Systems Engineering Management Plan (SEMP) is to describe the systems engineering approach and methods that will be integrated with established WHC engineering practices to enhance the WHC engineering management of the SNF Project. The scope of the SEMP encompasses the efforts needed to manage the WHC implementation of systems engineering on the SNF Project. This implementation applies to, and is tailored to the needs of the SNF project and all its subprojects, including all current and future subprojects

  14. Radioactive Air Emissions Notice of Construction (NOC) for the Fuel Removal Project

    International Nuclear Information System (INIS)

    JOHNSON, R.E.

    2000-01-01

    This NOC describes the activities to remove all spent nuclear fuel (SNF) assemblies from the spent fuel pool in the T Plant Complex 221-T canyon for interim storage in the Canister Storage Building (CSB). The unabated total effective dose equivalent (TEDE) estimated for the public hypothetical maximally exposed individual (MEI) is 5.7 E-6 millirem (mrem) per year for this fuel removal NOC. The abated TEDE conservatively is estimated to account for 2.9 E-9 mrem per year to the MEI

  15. Memorandum of Understanding Completion and Acceptance of the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    NISHIKAWA, L.D.

    1999-01-01

    This Memorandum of Understanding (MOU) is written to provide clear direction with respect to roles, responsibilities, obligations, and expectations of each organization identified. It functions as an agreement between the Operations, Construction Projects and Startup Organizations within the Spent Nuclear Fuels Project

  16. 76 FR 22668 - Shasta-Trinity National Forest; California; I-5 Corridor Fuels Reduction Project

    Science.gov (United States)

    2011-04-22

    ... environmental impact statement. SUMMARY: The Shasta Unit of the Shasta-Trinity National Forest is proposing a hazardous fuels treatment project to reduce the risk of life, property and resource values from a high... 20,025 acres of the project area is proposed for treatment. Treatment methods include prescribe fire...

  17. Average annual doses, lifetime doses and associated risk of cancer death for radiation workers in various fuel fabrication facilities in India

    International Nuclear Information System (INIS)

    Iyer, P.S.; Dhond, R.V.

    1980-01-01

    Lifetime doses based on average annual doses are estimated for radiation workers in various fuel fabrication facilities in India. For such cumulative doses, the risk of radiation-induced cancer death is computed. The methodology for arriving at these estimates and the assumptions made are discussed. Based on personnel monitoring records from 1966 to 1978, the average annual dose equivalent for radiation workers is estimated as 0.9 mSv (90 mrem), and the maximum risk of cancer death associated with this occupational dose as 1.35x10 -5 a -1 , as compared with the risk of death due to natural causes of 7x10 -4 a -1 and the risk of death due to background radiation alone of 1.5x10 -5 a -1 . (author)

  18. Dose evaluation model for radionuclides released from the spent nuclear fuel reprocessing plant in Rokkasho

    International Nuclear Information System (INIS)

    Hisamatsu, Shun'ichi; Iyogi, Takashi; Inaba, Jiro; Chiang, Jing-Hsien; Suwa, Hiroji; Koide, Mitsuo

    2007-01-01

    A dose evaluation model was developed for radionuclides released from the spent nuclear fuel reprocessing plant which is located in Rokkasho, Aomori Prefecture, and now undergoing test operation. The dose evaluation model suitable for medium- and long-term dose assessments for both prolonged and short-term releases of radionuclides to the atmosphere was developed on the PC. The ARAC-2, a particle tracing type dispersion model coupled with 3-D wind field calculation by a mass conservative model, was adopted as the atmospheric dispersion model. The terrestrial transfer model included movement in soil and groundwater as well as an agricultural and livestock farming system. The available site-specific social and environmental characteristics were incorporated in the model. Growing of the crops was also introduced and radionuclides absorbed were calculated from weight increase from the start of deposition to harvest, and transfer factors. Most of the computer code system of the models was completed by 2005, and this paper reports the results of the development. (author)

  19. Practical application of the dose limitation system in a uranium fuel fabrication plant

    International Nuclear Information System (INIS)

    Auricchio, S.; Cantoro, N.

    1982-01-01

    ICRP Publication 26 was published when the nuclear operators and the different national regulatory bodies were already in a position to understand the proposed dose limitation system and to apply it to nuclear activities. In Italy the basic principle of limiting individual risks and the search for increased protection were already applied in the radiation analysis of nuclear plants. These principles were applied during design (1972-74) and operation (1974-80) of the industrial fuel-element fabrication plant of the company Fabbricazioni Nucleari (F.N.) in Bosco Marengo. The paper reports on the criteria followed in the design stage, the organization and methods adopted for reducing the doses during operation, and the results achieved after a few years of plant activity. In view of the purely technical nature of this paper, the first principle of the dose limitation system (justification), which is more a political issue, is not taken into consideration; however, an assessment of the Italian context as at the end of the 1960s shows that the principle of justification of a practice was adequately taken into account when the construction of the F.N. plant was decided on. (author)

  20. Data base on dose reduction research projects for nuclear power plants: Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Khan, T.A.; Baum, J.W.

    1989-05-01

    This is the third volume in a series of reports that provide information on dose-reduction research and health physics technology for nuclear power plants. The information is taken from data base maintained by Brookhaven National Laboratory's ALARA Center for the Nuclear Regulatory Commission. This report presents information on 80 new projects, covering a wide area of activities. Projects on steam generator degradation, decontamination, robotics, improvement in reactor materials, and inspection techniques, among others, are described in the research section. The section on health physics technology includes some simple and very cost-effective projects to reduce radiation exposures. Collective dose data from the United States and other countries are also presented. In the conclusion, we suggest that although new advanced reactor design technology will eventually reduce radiation exposures at nuclear power plants to levels below serious concern, in the interim an aggressive approach to dose reduction remains necessary. 20 refs.

  1. A work bibliography on native food consumption, demography and lifestyle. Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Murray, C.E.; Lee, W.J.

    1992-12-01

    The purpose of this report is to provide a bibliography for the Native American tribe participants in the Hanford Environmental Dose Reconstruction (HEDR) Project to use. The HEDR Project`s primary objective is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy`s Hanford Site near Richland, Washington. Eight Native American tribes are responsible for estimating daily and seasonal consumption of traditional foods, demography, and other lifestyle factors that could have affected the radiation dose received by tribal members. This report provides a bibliography of recorded accounts that tribal researchers may use to verify their estimates. The bibliographic citations include references to information on the specific tribes, Columbia River plateau ethnobotany, infant feeding practices and milk consumption, nutritional studies and radiation, tribal economic and demographic characteristics (1940--1970), research methods, primary sources from the National Archives, regional archives, libraries, and museums.

  2. Fuels planning: science synthesis and integration; environmental consequences fact sheet 12: Water Erosion Prediction Project (WEPP) Fuel Management (FuMe) tool

    Science.gov (United States)

    William Elliot; David Hall

    2005-01-01

    The Water Erosion Prediction Project (WEPP) Fuel Management (FuMe) tool was developed to estimate sediment generated by fuel management activities. WEPP FuMe estimates sediment generated for 12 fuel-related conditions from a single input. This fact sheet identifies the intended users and uses, required inputs, what the model does, and tells the user how to obtain the...

  3. Demonstration project: Oxy-fuel combustion at Callide-A plant

    Energy Technology Data Exchange (ETDEWEB)

    Makino, Keiji; Misawa, Nobuhiro; Kiga, Takashi; Spero, Chris

    2007-07-01

    Oxy-fuel combustion is expected to be one of the promising systems on CO2 recovery from pulverized-coal power plant, and enable the CO2 to be captured in a more cost-effective manner compared to other CO2 recover process. An Australia-Japan consortium was established in 2004 specifically for the purpose of conducting a feasibility study on the application of oxy-fuel combustion to an existing pulverized-coal power plant that is Callide-A power plant No.4 unit at 30MWe owned by CS Energy in Australia. One of the important components in this study has been the recent comparative testing of three Australian coals under both oxy-fuel and air combustion conditions using the IHI combustion test facilities. The tests have yielded a number of important outcomes including a good comparison of normal air with oxy-fuel combustion, significant reduction in NOx mass emission rates under oxy-fuel combustion. On the basis of the feasibility study, the project under Australia-Japan consortium is now under way for applying oxy-fuel combustion to an existing plant by way of demonstration. In this project, a demonstration plant of oxy-fuel combustion will be completed by the end of 2008. This project aims at recovering CO2 from an actual power plant for storage. (auth)

  4. The PA projection of the clavicle: a dose-reducing technique.

    LENUS (Irish Health Repository)

    Mc Entee, Mark F

    2010-06-01

    This study compares dose and image quality during PA and AP radiography of the clavicle. The methodology involved a cadaver-based dose and image quality study. Results demonstrate a statistically significant 56.1 % (p dose with employment of PA and PA15 caudal projections. Reductions of 28.5 % (p doses to the eye were demonstrated for the PA. Differences in entrance-surface and exit doses were deemed non-significant. A 5.9 % (p dose reductions to the thyroid and breast when PA projection is chosen over the AP projection. The authors recommend the implementation of PA positioning for clavicle radiography.

  5. Acute Radiation Risk and BRYNTRN Organ Dose Projection Graphical User Interface

    Science.gov (United States)

    Cucinotta, Francis A.; Hu, Shaowen; Nounu, Hateni N.; Kim, Myung-Hee

    2011-01-01

    The integration of human space applications risk projection models of organ dose and acute radiation risk has been a key problem. NASA has developed an organ dose projection model using the BRYNTRN with SUM DOSE computer codes, and a probabilistic model of Acute Radiation Risk (ARR). The codes BRYNTRN and SUM DOSE are a Baryon transport code and an output data processing code, respectively. The risk projection models of organ doses and ARR take the output from BRYNTRN as an input to their calculations. With a graphical user interface (GUI) to handle input and output for BRYNTRN, the response models can be connected easily and correctly to BRYNTRN. A GUI for the ARR and BRYNTRN Organ Dose (ARRBOD) projection code provides seamless integration of input and output manipulations, which are required for operations of the ARRBOD modules. The ARRBOD GUI is intended for mission planners, radiation shield designers, space operations in the mission operations directorate (MOD), and space biophysics researchers. BRYNTRN code operation requires extensive input preparation. Only a graphical user interface (GUI) can handle input and output for BRYNTRN to the response models easily and correctly. The purpose of the GUI development for ARRBOD is to provide seamless integration of input and output manipulations for the operations of projection modules (BRYNTRN, SLMDOSE, and the ARR probabilistic response model) in assessing the acute risk and the organ doses of significant Solar Particle Events (SPEs). The assessment of astronauts radiation risk from SPE is in support of mission design and operational planning to manage radiation risks in future space missions. The ARRBOD GUI can identify the proper shielding solutions using the gender-specific organ dose assessments in order to avoid ARR symptoms, and to stay within the current NASA short-term dose limits. The quantified evaluation of ARR severities based on any given shielding configuration and a specified EVA or other mission

  6. Integrated Task Plans for the Hanford Environmental Dose Reconstruction Project, FY 1992 through May 1994

    International Nuclear Information System (INIS)

    Shipler, D.B.

    1992-09-01

    The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses from Hanford Site operations since 1944 to populations and individuals. The primary objective of work to be performed through May 1994 is to (1) determine the project's appropriate scope (space, time, radionuclides, pathways and individuals/population groups), (2) determine the project's appropriate level of accuracy (level of uncertainty in dose estimates) for the project, (3) complete model and data development, and (4) estimate doses for the Hanford Thyroid Disease Study (HTDS), representative individuals, and special populations as described herein. The plan for FY 1992 through May 1994 has been prepared based on activities and budgets approved by the Technical Steering Panel (TSP) at its meetings on August 19--20, 1991, and April 23--25, 1992. The activities can be divided into four broad categories: (1) model and data evaluation activities, (2)additional dose estimates, (3) model and data development activities, and (4)technical and communication support

  7. The Idaho Spent Fuel Project Update-January, 2003

    International Nuclear Information System (INIS)

    Roberts, R.; Tulberg, D.; Carter, C.

    2003-01-01

    The Department of Energy awarded a privatized contract to Foster Wheeler Environmental Corporation in May 2000 for the design, licensing, construction and operation of a spent nuclear fuel repackaging and storage facility. The Foster Wheeler Environmental Team consists of Foster Wheeler Environmental Corp. (the primary contractor), Alstec, RWE-Nukem, RIO Technical Services, Winston and Strawn, and Utility Engineering. The Idaho Spent Fuel (ISF) facility is an integral part of the DOE-EM approach to accelerating SNF disposition at the Idaho National Engineering and Environmental Laboratory (INEEL). Construction of this facility is also important in helping DOE to meet the provisions of the Idaho Settlement Agreement. The ISF Facility is a substantial facility with heavy shielding walls in the repackaging and storage bays and state-of-the-art features required to meet the provisions of 10 CFR 72 requirements. The facility is designed for a 40-year life

  8. SiC/SiC fuel cladding R and D Project 'SCARLET': Status and future plan

    International Nuclear Information System (INIS)

    Kishimoto, Hirotatsu; Kohyama, Akira

    2015-01-01

    This paper provides the recent progress in SiC/SiC development towards early utilisation for LWRs based on NITE method. After the March 11 Disaster in East-Japan, ensuring safe technology for LWR became a top priority R and D in nuclear energy policy of Japan. Along this line, replacement of Zircaloy claddings with SiC/SiC based fuel cladding is becoming one of the most attractive options and a MEXT fund based project, SCARLET, and a METI fund based project have been launched as 5-year termed projects at Muroran Institute of Technology. These projects care for NITE process for making long SiC/SiC fuel pins and connecting technology integration. The SCARLET project also includes coolant compatibility and irradiation effect evaluations as LWR and LMFBR materials. The outline and the present status of the SCARLET project will be briefly introduced in the present paper. (authors)

  9. BC Transit Fuel Cell Bus Project: Evaluation Results Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Post, M. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2014-02-01

    This report evaluates a fuel cell electric bus demonstration led by British Columbia Transit (BC Transit) in Whistler, Canada. BC Transit is collaborating with the California Air Resources Board and the U.S. Department of Energy's National Renewable Energy Laboratory to evaluate the buses in revenue service. This evaluation report covers two years of revenue service data on the buses from April 2011 through March 2013.

  10. Fuel Rod Consolidation Project: Phase 2, Final report: Volume 1

    International Nuclear Information System (INIS)

    1987-01-01

    This design report describes the NUS final design of the Prototype Spent Nuclear Fuel Rod Consolidation System. This summary presents the approach and the subsequent sections describe, in detail, the final design. Detailed data, drawings, and the design Basis Accident Report are provided in Volumes II thru V. The design as presented, represents one cell of a multicell facility for the dry consolidation of any type of PWR and BWR fuel used in the United States LWR industry that will exceed 1% of the fuel inventory at the year 2000. The system contains the automatically-controlled equipment required to consolidate 750MT (heavy metal)/year, at 75% availability. The equipment is designed as replaceable components using state-of-the-art tchnology. The control system utilizes the most advanced commercially available equipment on the market today. Two state-of-the-art advanced servo manipulators are provided for system maintenance. In general the equipment is designed utilizing fabricated and commercial components. For example, the main drive systems use commercially available roller screws. These rollers screws have 60,000 hours of operation in nuclear power plants and have been used extensively in other applications. The motors selected represent the most advanced designed servo motors on the market today for the precision control of machinery. In areas where precise positioning was not required, less expensive TRW Globe motors were selected. These are small compact motors with a long history of operations in radiation environments. The Robotic Bridge Transporters are modified versions of existing bridge cranes for remote automatic operations. Other equipment such as the welder for fuel canister closure operations is a commercially available product with an operating history applicable to this process. In general, this approach was followed throughout the design of all the equipment and will enable the system to be developed without costly development programs

  11. Ultra-clean Fischer-Tropsch (F-T) Fuels Production and Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Stephen P. Bergin

    2006-06-30

    The objective of the DOE-NETL Fischer-Tropsch (F-T) Production and Demonstration Program was to produce and evaluate F-T fuel derived from domestic natural gas. The project had two primary phases: (1) fuel production of ultra-clean diesel transportation fuels from domestic fossil resources; and (2) demonstration and performance testing of these fuels in engines. The project also included a well-to-wheels economic analysis and a feasibility study of small-footprint F-T plants (SFPs) for remote locations such as rural Alaska. During the fuel production phase, ICRC partnered and cost-shared with Syntroleum Corporation to complete the mechanical design, construction, and operation of a modular SFP that converts natural gas, via F-T and hydro-processing reactions, into hydrogensaturated diesel fuel. Construction of the Tulsa, Oklahoma plant started in August 2002 and culminated in the production of over 100,000 gallons of F-T diesel fuel (S-2) through 2004, specifically for this project. That fuel formed the basis of extensive demonstrations and evaluations that followed. The ultra-clean F-T fuels produced had virtually no sulfur (less than 1 ppm) and were of the highest quality in terms of ignition quality, saturation content, backend volatility, etc. Lubricity concerns were investigated to verify that commercially available lubricity additive treatment would be adequate to protect fuel injection system components. In the fuel demonstration and testing phase, two separate bus fleets were utilized. The Washington DC Metropolitan Area Transit Authority (WMATA) and Denali National Park bus fleets were used because they represented nearly opposite ends of several spectra, including: climate, topography, engine load factor, mean distance between stops, and composition of normally used conventional diesel fuel. Fuel evaluations in addition to bus fleet demonstrations included: bus fleet emission measurements; F-T fuel cold weather performance; controlled engine dynamometer

  12. FY 1991 Task plans for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    1991-04-01

    The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses from Hanford Site operations since 1944 to populations and individuals. The objectives of work in Fiscal Year (FY) 1991 are to analyze data and models used in Phase 1 and restructure the models to increase accuracy and reduce uncertainty in dose estimation capability. Databases will be expanded and efforts will begin to determine the appropriate scope (space, time, radionuclides, pathways and individuals/population groups) and accuracy (level of uncertainty in dose estimates) for the project. Project scope and accuracy requirements, once defined, can be translated into additional model and data requirements later in the project. Task plans for FY 1991 have been prepared based on activities approved by the Technical Steering Panel (TSP) in October 1990 and mid-year revisions discussed at the TSP planning/budget workshop in February 1991. The activities can be divided into two broad categories: (1) model and data development and evaluation, (2) project, technical and communication support. 3 figs., 1 tab

  13. European project for developing general guidelines for harmonising internal dose assessment procedures (IDEAS)

    International Nuclear Information System (INIS)

    Andrasi, A.; Bailey, M.; Puncher, M.; Berkovski, V.; Eric Blanchardon, E.; Jourdain, J.-R.; Carlo-Maria Castellani, C.-M.; Doerfel, H.; Christian Hurtgen, Ch.; Le Guen, B.

    2003-01-01

    Several international intercomparison exercises on intake and internal dose assessments from monitoring data led to the conclusion that the results calculated by different participants varied significantly mainly because of the wide variety of methods and assumptions applied in the assessment procedure. Based on these experiences the need for harmonisation of the procedures has been formulated as an EU research project under the 5 th Framework Programme (2001-2005), with the aim of developing general guidelines for standardising assessments of intakes and internal doses. In the IDEAS project eight institutions from seven European countries are participating using inputs also from internal dosimetry professionals from across Europe to ensure broad consensus in the outcome of the project. The IDEAS project is explained

  14. Global Threat Reduction Initiative Fuel Thermo-Physical Characterization Project: Sample Management Plan

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pereira, Mario M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Steen, Franciska H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-01-01

    This sample management plan provides guidelines for sectioning, preparation, acceptance criteria, analytical path, and end-of-life disposal for the fuel element segments utilized in the Global Threat Reduction Initiative (GTRI), Fuel Thermo-Physical Characterization Project. The Fuel Thermo-Physical Characterization Project is tasked with analysis of irradiated Low Enriched Uranium (LEU) Molybdenum (U-Mo) fuel element samples to support the GTRI conversion program. Sample analysis may include optical microscopy (OM), scanning electron microscopy (SEM) fuel-surface interface analysis, gas pycnometry (density) measurements, laser flash analysis (LFA), differential scanning calorimetry (DSC), thermogravimetry and differential thermal analysis with mass spectroscopy (TG /DTA-MS), Inductively Coupled Plasma Spectrophotometry (ICP), alpha spectroscopy, and Thermal Ionization Mass Spectroscopy (TIMS). The project will utilize existing Radiochemical Processing Laboratory (RPL) operating, technical, and administrative procedures for sample receipt, processing, and analyses. Test instructions (TIs), which are documents used to provide specific details regarding the implementation of an existing RPL approved technical or operational procedure, will also be used to communicate to staff project specific parameters requested by the Principal Investigator (PI). TIs will be developed, reviewed, and issued in accordance with the latest revision of the RPL-PLN-700, RPL Operations Plan. Additionally, the PI must approve all project test instructions and red-line changes to test instructions.

  15. Irradiation project of SiC/SiC fuel pin 'INSPIRE': Status and future plan

    International Nuclear Information System (INIS)

    Kohyama, Akira; Kishimoto, Hirotatsu

    2015-01-01

    After the March 11 Disaster in East-Japan, Research and Development towards Ensuring Nuclear Safety Enhancement for LWR becomes a top priority R and D in nuclear energy policy of Japan. The role of high temperature non-metallic materials, such as SiC/SiC, is becoming important for the advanced nuclear reactor systems. SiC fibre reinforced SiC composite has been recognised to be the most attractive option for the future, now, METI fund based project, INSPIRE, has been launched as 5-year termed project at OASIS in Muroran Institute of Technology aiming at early realisation of this system. INSPIRE is the irradiation project of SiC/SiC fuel pins aiming to accumulate material, thermal, irradiation effect data of NITE-SiC/SiC in BWR environment. Nuclear fuel inserted SiC/SiC fuel pins are planned to be installed in the Halden reactor. The project includes preparing the NITE-SiC/SiC tubes, joining of end caps, preparation of rigs to control the irradiation environment to BWR condition and the instruments to measure the condition of rigs and pins in operation. Also, basic neutron irradiation data will be accumulated by SiC/SiC coupon samples currently under irradiation in BR2. The output from this project may present the potentiality of NITE-SiC/SiC fuel cladding with the first stage fuel-cladding interaction. (authors)

  16. IAEA TC Project 'Strengthening safety and reliability of fuel and materials in nuclear power plants'

    International Nuclear Information System (INIS)

    Makihara, Y.

    2008-01-01

    The Regional TC Project in Europe RER9076 'Strengthening Safety and Reliability of Fuel and Materials in Nuclear Power Plants' was launched in 2003 as a four-year project and was subsequently extended in 2006 to run through 2008. The purpose of the Project is to support the Central and Eastern European countries with the necessary tools to fulfill their own fuel and material licensing needs. The main objective will be to provide quality data on fuel and materials irradiated in power reactors and in dedicated experiments carried out in material test reactors (MTRs). Within the framework of the Project, ten tasks were implemented. These included experiments performed at the test facilities in the region, training courses and workshops related to fuel safety. While several tasks are expected to be completed by the end of RER9076, some remain. It would be desirable to initiate a new RER Project from the next TC cycle (2009-2011) in order to take over RER9076 and to implement new tasks required for enhancing fuel safety in the region. (author)

  17. Soil Remediation Demonstration Project: Biodegradation of Heavy Fuel Oils

    National Research Council Canada - National Science Library

    Reynolds, Charles

    1997-01-01

    .... Low-cost treatments applicable to small-scale spills are needed. The object of this CPAR project was to examine using cost-effective, on-site bioremediation techniques for heavy-oil-contaminated soil in cold regions...

  18. Production of wood derived fuels. Review of research projects; Puupolttoaineiden tuotantotekniikka. Tutkimusalueen katsaus

    Energy Technology Data Exchange (ETDEWEB)

    Korpilahti, A [Metsaeteho Oy, Helsinki (Finland)

    1997-12-01

    The research and development work was very active on the area of wood derived fuels during the past year 1996. Totally some 40 projects were going on, and till the end of the year about 15 projects were completed. The projects broadly covered the research area focusing from material flows, productivity studies, basic wood properties to several case studies. When new production methods and machinery was introduced earlier by demonstration projects, now they were investigated by follow up projects. The economical and quality results of logging residue harvesting and comminution seem quite satisfactory, but integrated methods and production chains still need research and development. (orig.)

  19. The Effect of Fuel Dose Division on The Emission of Toxic Components in The Car Diesel Engine Exhaust Gas

    Directory of Open Access Journals (Sweden)

    Pietras Dariusz

    2016-09-01

    Full Text Available The article discusses the effect of fuel dose division in the Diesel engine on smoke opacity and composition of the emitted exhaust gas. The research activities reported in the article include experimental examination of a small Diesel engine with Common Rail type supply system. The tests were performed on the engine test bed equipped with an automatic data acquisition system which recorded all basic operating and control parameters of the engine, and smoke opacity and composition of the exhaust gas. The parameters measured during the engine tests also included the indicated pressure and the acoustic pressure. The tests were performed following the pre-established procedure in which 9 engine operation points were defined for three rotational speeds: 1500, 2500 and 3500 rpm, and three load levels: 25, 40 and 75 Nm. At each point, the measurements were performed for 7 different forms of fuel dose injection, which were: the undivided dose, the dose divided into two or three parts, and three different injection advance angles for the undivided dose and that divided into two parts. The discussion of the obtained results includes graphical presentation of contests of hydrocarbons, carbon oxide, and nitrogen oxides in the exhaust gas, and its smoke opacity. The presented analyses referred to two selected cases, out of nine examined engine operation points. In these cases the fuel dose was divided into three parts and injected at the factory set control parameters. The examination has revealed a significant effect of fuel dose division on the engine efficiency, and on the smoke opacity and composition of the exhaust gas, in particular the content of nitrogen oxides. Within the range of low loads and rotational speeds, dividing the fuel dose into three parts clearly improves the overall engine efficiency and significantly decreases the concentration of nitrogen oxides in the exhaust gas. Moreover, it slightly decreases the contents of hydrocarbons and

  20. Procedure for matching synfuel users with potential suppliers. Appendix B. Proposed and ongoing synthetic fuel production projects

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-08-07

    To assist the Department of Energy, Office of Fuels Conversion (OFC), in implementing the synthetic fuel exemption under the Powerplant and Industrial Fuel Use Act (FUA) of 1978, Resource Consulting Group, Inc. (RCG), has developed a procedure for matching prospective users and producers of synthetic fuel. The matching procedure, which involves a hierarchical screening process, is designed to assist OFC in: locating a supplier for a firm that wishes to obtain a synthetic fuel exemption; determining whether the fuel supplier proposed by a petitioner is technically and economically capable of meeting the petitioner's needs; and assisting the Synthetic Fuels Corporation or a synthetic fuel supplier in evaluating potential markets for synthetic fuel production. A data base is provided in this appendix on proposed and ongoing synthetic fuel production projects to be used in applying the screening procedure. The data base encompasses a total of 212 projects in the seven production technologies.

  1. Dose calculation for accident situations at WWR-S type spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Margeanu, S.; Florescu, G.

    2006-01-01

    Full text: The Spent Nuclear Fuel Repository at IFIN-HH Bucharest (SNFR IFIN-HH) consists in four pools, repository hall, radiological monitoring system, ventilation system and auxiliary systems. At the moment the remaining activity in the repository is about 3500 Ci. Despite of the small activity, for emergency preparedness purposes, several accident scenarios, with a non zero probability of occurrence during the repository lifetime, have been postulated. Evaluations of radiological consequences to personnel, general public and environment, for each accident scenario have been performed. The radioactive inventory was evaluated with ORIGEN code from SCALE computer code system and radiological consequences were evaluated with COSYMA computer code. Assumptions for the source term determination, meteorological conditions and release, are presented. The calculated values of doses and risk are also presented. The impact of these accident scenarios on population and environment is also discussed. (authors)

  2. Hydrogen Research for Spaceport and Space-Based Applications: Fuel Cell Projects

    Science.gov (United States)

    Anderson, Tim; Balaban, Canan

    2008-01-01

    The activities presented are a broad based approach to advancing key hydrogen related technologies in areas such as fuel cells, hydrogen production, and distributed sensors for hydrogen-leak detection, laser instrumentation for hydrogen-leak detection, and cryogenic transport and storage. Presented are the results from research projects, education and outreach activities, system and trade studies. The work will aid in advancing the state-of-the-art for several critical technologies related to the implementation of a hydrogen infrastructure. Activities conducted are relevant to a number of propulsion and power systems for terrestrial, aeronautics and aerospace applications. Fuel cell research focused on proton exchange membranes (PEM), solid oxide fuel cells (SOFC). Specific technologies included aircraft fuel cell reformers, new and improved electrodes, electrolytes, interconnect, and seals, modeling of fuel cells including CFD coupled with impedance spectroscopy. Research was conducted on new materials and designs for fuel cells, along with using embedded sensors with power management electronics to improve the power density delivered by fuel cells. Fuel cell applications considered were in-space operations, aviation, and ground-based fuel cells such as; powering auxiliary power units (APUs) in aircraft; high power density, long duration power supplies for interplanetary missions (space science probes and planetary rovers); regenerative capabilities for high altitude aircraft; and power supplies for reusable launch vehicles.

  3. Dose reduction in chest CT: Comparison of the adaptive iterative dose reduction 3D, adaptive iterative dose reduction, and filtered back projection reconstruction techniques

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Yoshitake, E-mail: yamada@rad.med.keio.ac.jp [Department of Diagnostic Radiology, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan); Jinzaki, Masahiro, E-mail: jinzaki@rad.med.keio.ac.jp [Department of Diagnostic Radiology, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan); Hosokawa, Takahiro, E-mail: hosokawa@rad.med.keio.ac.jp [Department of Diagnostic Radiology, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan); Tanami, Yutaka, E-mail: tanami@rad.med.keio.ac.jp [Department of Diagnostic Radiology, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan); Sugiura, Hiroaki, E-mail: hsugiura@rad.med.keio.ac.jp [Department of Diagnostic Radiology, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan); Abe, Takayuki, E-mail: tabe@z5.keio.jp [Center for Clinical Research, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan); Kuribayashi, Sachio, E-mail: skuribay@a5.keio.jp [Department of Diagnostic Radiology, Keio University School of Medicine, 35 Shinanomachi, Shinjuku-ku, Tokyo 160-8582 (Japan)

    2012-12-15

    Objectives: To assess the effectiveness of adaptive iterative dose reduction (AIDR) and AIDR 3D in improving the image quality in low-dose chest CT (LDCT). Materials and methods: Fifty patients underwent standard-dose chest CT (SDCT) and LDCT simultaneously, performed under automatic exposure control with noise index of 19 and 38 (for a 2-mm slice thickness), respectively. The SDCT images were reconstructed with filtered back projection (SDCT-FBP images), and the LDCT images with FBP, AIDR and AIDR 3D (LDCT-FBP, LDCT-AIDR and LDCT-AIDR 3D images, respectively). On all the 200 lung and 200 mediastinal image series, objective image noise and signal-to-noise ratio (SNR) were measured in several regions, and two blinded radiologists independently assessed the subjective image quality. Wilcoxon's signed rank sum test with Bonferroni's correction was used for the statistical analyses. Results: The mean dose reduction in LDCT was 64.2% as compared with the dose in SDCT. LDCT-AIDR 3D images showed significantly reduced objective noise and significantly increased SNR in all regions as compared to the SDCT-FBP, LDCT-FBP and LDCT-AIDR images (all, P ≤ 0.003). In all assessments of the image quality, LDCT-AIDR 3D images were superior to LDCT-AIDR and LDCT-FBP images. The overall diagnostic acceptability of both the lung and mediastinal LDCT-AIDR 3D images was comparable to that of the lung and mediastinal SDCT-FBP images. Conclusions: AIDR 3D is superior to AIDR. Intra-individual comparisons between SDCT and LDCT suggest that AIDR 3D allows a 64.2% reduction of the radiation dose as compared to SDCT, by substantially reducing the objective image noise and increasing the SNR, while maintaining the overall diagnostic acceptability.

  4. BC Transit Fuel Cell Bus Project Evaluation Results: Second Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Post, M.

    2014-09-01

    Second report evaluating a fuel cell electric bus (FCEB) demonstration led by British Columbia Transit (BC Transit) in Whistler, Canada. BC Transit is collaborating with the California Air Resources Board and the U.S. Department of Energy's National Renewable Energy Laboratory to evaluate the buses in revenue service. NREL published its first report on the demonstration in February 2014. This report is an update to the previous report; it covers 3 full years of revenue service data on the buses from April 2011 through March 2014 and focuses on the final experiences and lessons learned.

  5. Fuel elements assembling for the DON project exponential experience

    International Nuclear Information System (INIS)

    Anca Abati, R. de

    1966-01-01

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs

  6. Current status of the tasks performed by the Nuclear Fuel Department for the CAREM project

    International Nuclear Information System (INIS)

    Kaufmann, Federico; Perez, Lidia E.; Perez, Aaldo; Marchi, Daniel E.

    2009-01-01

    CAREM Project required to the Nuclear Fuels Department to perform the necessary tasks to define the powder characteristics and natural UO 2 -Gd 2 O 3 , UO 2 -1.8 and 3.1 % enrichment fuel pellets manufacturing and control parameters. To start with these tasks it was necessary to hire and train staff, begin the licensing process of the facilities, update infrastructure works and equipment of existing facilities. A brief description of the tasks performed is shown. (author)

  7. Reduction of fuel enrichment for research reactors built-up in accordance with Russian (Soviet) projects

    International Nuclear Information System (INIS)

    Aleksandrov, A.B.; Enin, A.A.; Tkachyov, A.A.

    2001-01-01

    In accordance with the Russian program of reduced enrichment for research and test reactors (RERTR) built-up in accordance with Russian (Soviet) projects, AO 'NCCP' performs works on FA fabrication with reduced enrichment fuel. The main trends and results of performed works on research reactors FEs and FAs based on UO 2 and U-9%Mo fuel with U 235 19.7% enrichment are described. (author)

  8. Coming on stream: Financing biomass and alternative-fuel projects in the 1990s

    International Nuclear Information System (INIS)

    Mumford, E.B. Jr.

    1993-01-01

    Biomass-energy and alternative-fuels projects make environmental sense, but do they make economic sense? In the current project-finance environment, moving ideas off the drawing board and transforming them into reality takes more than vision and commitment; it takes the ability to understand and address the financial markets' perception of risk. This paper examines the state of the project-finance market, both as it pertains to biomass and alternative-fuels projects and in more general terms, focusing on what project sponsors and developers need to dot to obtain both early-state and construction/term financing, and the role a financial adviser can play in helping ensure access to funds at all stages

  9. Assessment of axial gamma dose rate profile on irradiated fuel assembly using polycarbonate film and perspex dosimeters

    International Nuclear Information System (INIS)

    Joshi, V.B.; Janardhanan, S.; Pillai, P.R.; Somanathan, K.; Narayan, K.K.; John, J.; Kutty, K.N.; Deo, V.R.; Popli, O.L.

    1986-01-01

    The dose-rate profile of irradiated fuel rod is required for optimisation of radiation shielding from safety point of view during storage, handling and metallurgical examination. Since the dose-rates are in kilogray per hour, their determination requires special evaluation techniques. This paper illustrates the application of Makrofol-N and red perspex (AERE 4034B) for this purpose. They are compared with CaSO 4 :Dy thermoluminescence dosimeter. (author). 4 refs

  10. The IAEA's international project on innovative nuclear reactors and fuel cycles (INPRO)

    International Nuclear Information System (INIS)

    Kuptiz, Juergen; )

    2002-01-01

    This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited all interested Member states to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology and invited Member states to consider to contribute to a task force on innovative nuclear reactors and fuel cycle

  11. Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gardner, Levi D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huber, Tanja K. [Technische Universität München, Munich (Germany); Breitkreutz, Harald [Technische Universität München, Munich (Germany)

    2015-02-11

    The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

  12. Handbook of selected organ doses for projections common in pediatric radiology

    International Nuclear Information System (INIS)

    Rosenstein, M.; Beck, T.J.; Warner, G.G.

    1979-05-01

    This handbook contains data from which absorbed dose (mrad) to selected organs can be estimated for common projections in pediatric radiology. The organ doses are for three reference patients: a newborn (0 to 6 months), a 1-year old child, and a 5-year old child. One intent of the handbook is to permit the user to evaluate the effect on organ dose to these reference pediatric patients as a function of certain changes in technical parameters used in or among facilities. A second intent is to permit a comparison to be made of organ doses as a function of age. This comparison can be extended to a reference adult by referring to the previous Handbook of Selected Organ Doses fo Projections Common in Diagnostic Radiology, FDA 76-8031. Assignment of organ doses to individual pediatric patients using the Handbook data is not recommended unless the physical characteristics of the patient closely correlate with one of the three reference pediatric patients given in Appendix A

  13. Draft Air Pathway Report: Phase 1 of the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-20

    This report summarizes the air pathway portion of the first phase of the Hanford Environmental Dose Reconstruction (HEDR) Project, conducted by Battelle staff at the Pacific Northwest Laboratory under the direction of an independent Technical Steering Panel. The HEDR Project is estimating historical radiation doses that could have been received by populations near the Department of Energy's Hanford Site, in southeastern Washington State. Phase 1 of the air-pathway dose reconstruction sought to determine whether dose estimates could be calculated for populations in the 10 counties nearest the Hanford Site from atmospheric releases of iodine-131 from the site from 1944--1947. Phase 1 demonstrated the following: HEDR-calculated source-term estimates of iodine-131 releases to the atmosphere were within 20% of previously published estimates; calculated vegetation concentrations of iodine-131 agree well with previously published measurements; the highest of the Phase 1 preliminary dose estimates to the thyroid are consistent with independent, previously published estimates of doses to maximally exposed individuals; and relatively crude, previously published measurements of thyroid burdens for Hanford workers are in the range of average burdens that the HEDR model estimated for similar reference individuals'' for the period 1944--1947. 4 refs., 10 figs., 9 tabs.

  14. Application of the BISON Fuel Performance Code of the FUMEX-III Coordinated Research Project

    International Nuclear Information System (INIS)

    Williamson, R.L.; Novascone, S.R.

    2013-01-01

    Since 1981, the International Atomic Energy Agency (IAEA) has sponsored a series of Coordinated Research Projects (CRP) in the area of nuclear fuel modeling. These projects have typically lasted 3-5 years and have had broad international participation. The objectives of the projects have been to assess the maturity and predictive capability of fuel performance codes, support interaction and information exchange between countries with code development and application needs, build a database of well- defined experiments suitable for code validation, transfer a mature fuel modeling code to developing countries, and provide guidelines for code quality assurance and code application to fuel licensing. The fourth and latest of these projects, known as FUMEX-III1 (FUel Modeling at EXtended Burnup- III), began in 2008 and ended in December of 2011. FUMEX-III was the first of this series of fuel modeling CRP's in which the INL participated. Participants met at the beginning of the project to discuss and select a set of experiments ('priority cases') for consideration during the project. These priority cases were of broad interest to the participants and included reasonably well-documented and reliable data. A meeting was held midway through the project for participants to present and discuss progress on modeling the priority cases. A final meeting was held at close of the project to present and discuss final results and provide input for a final report. Also in 2008, the INL initiated development of a new multidimensional (2D and 3D) multiphysics nuclear fuel performance code called BISON, with code development progressing steadily during the three-year FUMEX-III project. Interactions with international fuel modeling researchers via FUMEX-III played a significant role in the BISON evolution, particularly influencing the selection of material and behavioral models which are now included in the code. The FUMEX-III cases are generally integral fuel rod experiments occurring

  15. Project development laboratories energy fuels and oils based on NRU “MPEI”

    Science.gov (United States)

    Burakov, I. A.; Burakov, A. Y.; Nikitina, I. S.; Khomenkov, A. M.; Paramonova, A. O.; Khtoo Naing, Aung

    2017-11-01

    In the process of improving the efficiency of power plants a hot topic is the use of high-quality fuels and lubricants. In the process of transportation, preparation for use, storage and maintenance of the properties of fuels and lubricants may deteriorate, which entails a reduction in the efficiency of power plants. One of the ways to prevent the deterioration of the properties is a timely analysis of the relevant laboratories. In this day, the existence of laboratories of energy fuels and energy laboratory oil at thermal power stations is satisfactory character. However, the training of qualified personnel to work in these laboratories is a serious problem, as the lack of opportunities in these laboratories a complete list of required tests. The solution to this problem is to explore the possibility of application of methods of analysis of the properties of fuels and lubricants in the stage of training and re-training of qualified personnel. In this regard, on the basis of MPEI developed laboratory projects of solid, liquid and gaseous fuels, power and energy oils and lubricants. Projects allow for a complete list of tests required for the timely control of properties and prevent the deterioration of these properties. Assess the financial component of the implementation of the developed projects based on the use of modern equipment used for tests. Projects allow for a complete list of tests required for the timely control of properties and prevent the deterioration of these properties.

  16. Overview of Graphical User Interface for ARRBOD (Acute Radiation Risk and BRYNTRN Organ Dose Projection)

    Science.gov (United States)

    Kim, Myung-Hee Y.; Hu, Shaowen; Nounu, Hatem N.; Cucinotta, Francis A.

    2010-01-01

    Solar particle events (SPEs) pose the risk of acute radiation sickness (ARS) to astronauts, because organ doses from large SPEs may reach critical levels during extra vehicular activities (EVAs) or lightly shielded spacecraft. NASA has developed an organ dose projection model of Baryon transport code (BRYNTRN) with an output data processing module of SUMDOSE, and a probabilistic model of acute radiation risk (ARR). BRYNTRN code operation requires extensive input preparation, and the risk projection models of organ doses and ARR take the output from BRYNTRN as an input to their calculations. With a graphical user interface (GUI) to handle input and output for BRYNTRN, these response models can be connected easily and correctly to BRYNTRN in a user friendly way. The GUI for the Acute Radiation Risk and BRYNTRN Organ Dose (ARRBOD) projection code provides seamless integration of input and output manipulations required for operations of the ARRBOD modules: BRYNTRN, SUMDOSE, and the ARR probabilistic response model. The ARRBOD GUI is intended for mission planners, radiation shield designers, space operations in the mission operations directorate (MOD), and space biophysics researchers. Assessment of astronauts organ doses and ARS from the exposure to historically large SPEs is in support of mission design and operation planning to avoid ARS and stay within the current NASA short-term dose limits. The ARRBOD GUI will serve as a proof-of-concept for future integration of other risk projection models for human space applications. We present an overview of the ARRBOD GUI product, which is a new self-contained product, for the major components of the overall system, subsystem interconnections, and external interfaces.

  17. Radiation doses due to natural radon gas releases from the final disposal facility of spent fuel

    International Nuclear Information System (INIS)

    Vesterbacka, K.; Arvela, H.

    1998-03-01

    Building an underground repository for the spent nuclear fuel increases releases of natural radon gas. In the report the radon releases, the resulting doses as well as the radon concentration in the repository air are investigated. There are four optional building locations for the underground repository and three different strategies of construction. Optional sites are Olkiluoto of Eurajoki, Romuvaara of Kuhmo, Haestholmen of Loviisa and Kivetty of Aeaenekoski. The most significant radon sources in the underground repository are the rockwalls and the groundwater leaking to the repository. High groundwater radon concentrations can increase significantly radon concentration in the repository air despite the groundwater leak rate is low. The radon source strength from the rockwalls, groundwater and macadam spreaded on the floor of the repository is estimated in this report. Using these results the radon concentration in the repository is calculated for several air exchange rates. Data from petrological studies performed at the optional building sites as well as the measurement data of the Radiation and Nuclear Safety Authority has been utilized. Rough approximations were needed when estimating the radon source strength. The estimated total radon source strength varies between 1 - 600 MBq/h depending on the repository construction strategy. Repository indoor air radon concentration with no air exchange varies between 0,7 - 120 kBq/m 3 . Using the most probable estimates on radon source strength, the allowed indoor radon concentration of 400 Bq/m 3 at workplaces is achieved by using the air exchange rate of 0,5 l/h in every optional repository. Repository exhaust air and the pile of macadam increases the radon levels in the environment. The radiation dose to the critical person depends on the open volume of the repository. The annual radiation dose calculated from the most probable radon source strength at the distance of 500 metres is below 0,005 mSv at all sites

  18. Incorporating detrital conditioning in outdoor microcosms dosed with JP-8 jet fuel

    International Nuclear Information System (INIS)

    Matthews, R.A.; Markiewicz, A.; Harter, V.; Landis, W.G.

    1995-01-01

    The authors have developed an outdoor microcosm system that incorporates detrital conditioning to test the hypothesis that microbiota play a critical role in altering the community response to hydrocarbon toxicants. The microcosms were constructed using 568 L tanks, arranged in 6 units of 4 tanks, with each unit equidistant from a central conditions tank (CT). During pre-treatment, the microcosms and CT were filled with nutrient-amended well water, artificial sediment, leaf packs containing dried maple leaves, elodea fragment, and unglazed tiles for periphyton growth. Water circulation was maintained at the rate of 24 exchanges per day. After four weeks, invertebrates from local ponds were added to the CT. Leafpacks were added to the CT and microcosms every two weeks; eight week old packs were discarded after returning invertebrates to the CT. On a weekly basis, 25% of the sediments, leaf packs, tiles, and elodea from each microcosms were transferred to another microcosm; the CT walls and tiles were scraped; an the water quality was monitored. Circulation was discontinued one week prior to dosing. On 4/12/96, the microcosms were dosed to contain 0--0.25 microg/L of JP-8 jet fuel. Within two weeks the GC/MS hydrocarbon concentrations were very low in the water column of the highest treatment group. There has been little acute toxicity, despite selecting doses that caused severe, acute toxicity in laboratory microcosm studies. The presence of a complex, detritus-based microbial community appears to mitigate the influences of the toxicant on the microcosms

  19. Final report on LDRD project ''proliferation-resistant fuel cycles''

    International Nuclear Information System (INIS)

    Brown, N W; Hassberger, J A.

    1999-01-01

    This report provides a summary of LDRD work completed during 1997 and 1998 to develop the ideas and concepts that lead to the Secure, Transportable, Autonomous Reactor (STAR) program proposals to the DOE Nuclear Energy Research Initiative (NERI). The STAR program consists of a team of three national laboratories (LLNL, ANL, and LANL), three universities, (UC Berkeley, TAMU, and MIT) and the Westinghouse Research Center. Based on the LLNL work and their own efforts on related work this team prepared and integrated a package of twelve proposals that will carry the LDRD work outlined here into the next phase of development. We are proposing to develop a new nuclear system that meets stringent requirements for a high degree of safety and proliferation resistance, and also deals directly with the related nuclear waste and spent fuel management issues

  20. Projections of spent fuel to be discharged by the U.S. nuclear power industry

    International Nuclear Information System (INIS)

    Alexander, C.W.; Kee, C.W.; Croff, A.G.; Blomeke, J.O.

    1977-10-01

    Calculated properties of spent fuel projected to be discharged and accumulated by the U.S. nuclear power industry through the year 2031 A.D. are presented. The projections are based on installed nuclear capacities of 380 and 543 GW(e) in the year 2000 and 2030, respectively. They include compilations of the grams of the elements, curies of radioactivity, thermal decay power, photon and neutron emission rates, and radiotoxicities of the assemblies that are accumulated at a Spent Unreprocessed Fuel Facility (SURFF), allowing for delays of 5 and 10 years before shipment to SURFF

  1. An overview of the fuels and materials testing programme at the OECD Halden Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-08-01

    The fuels and materials testing programme of the OECD Halden Reactor Project is aimed at investigations of fuel and cladding properties at high burnup, water chemistry effects and in-core materials ageing problems. For the execution of this programme, different types of irradiation rigs and experimental facilities providing typical power reactors conditions are available. Data are obtained from in-core sensors developed at the Halden Project; these are shortly described. An overview of the current test programme and the scope of the following years are briefly presented. (author). 5 refs, 3 figs.

  2. Spent Nuclear Fuel Project technical baseline document. Fiscal year 1995: Volume 1, Baseline description

    International Nuclear Information System (INIS)

    Womack, J.C.; Cramond, R.; Paedon, R.J.

    1995-01-01

    This document is a revision to WHC-SD-SNF-SD-002, and is issued to support the individual projects that make up the Spent Nuclear Fuel Project in the lower-tier functions, requirements, interfaces, and technical baseline items. It presents results of engineering analyses since Sept. 1994. The mission of the SNFP on the Hanford site is to provide safety, economic, environmentally sound management of Hanford SNF in a manner that stages it to final disposition. This particularly involves K Basin fuel, although other SNF is involved also

  3. Guide for subdivision of spent fuel pool. Project UNESA MAAP5-SFP

    International Nuclear Information System (INIS)

    Martinez Barrios, M.; Garcia Gonzalez, M.; Perez Martin, F. J.

    2013-01-01

    The main goal of the UNESA MAAP5-SFP project is to analyze the capabilities of MAAP5 code and, particularly, the Spent Fuel Pool (SFP) module in order to tackle its modeling and facilitate the development of specific SFP models of Spanish NPPs. Within the project, Empresarios Agrupados (EEAA) is the responsible for the development of the Guide for the subdivision of the Spent Fuel Pool (SFP). This Guide includes a theoretical description of the model that is used by the code and a sequence of practical cases with the aim to evaluate the influence of specific parameters

  4. Projections of spent fuel to be discharged by the U. S. nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, C.W.; Kee, C.W.; Croff, A.G.; Blomeke, J.O.

    1977-10-01

    Calculated properties of spent fuel projected to be discharged and accumulated by the U.S. nuclear power industry through the year 2031 A.D. are presented. The projections are based on installed nuclear capacities of 380 and 543 GW(e) in the year 2000 and 2030, respectively. They include compilations of the grams of the elements, curies of radioactivity, thermal decay power, photon and neutron emission rates, and radiotoxicities of the assemblies that are accumulated at a Spent Unreprocessed Fuel Facility (SURFF), allowing for delays of 5 and 10 years before shipment to SURFF.

  5. Financial investments in fuel cells and hydrogen projects in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Brito de Matos, Maiana; Neves, Newton Pimenta Jr.; Silva, Ennio Peres da; Silva Pinto, Cristiano [Universidade Estadual de Campinas (UNICAMP), SP (Brazil)

    2010-07-01

    This work aims to identify, classify and account for the investments in hydrogen and fuel cells from 1999 to 2007 made by the public and private sectors in Brazil. Two methodologies were applied to obtain the data for this study. The Top-Down methodology was used to obtain the information from the sponsoring agencies, institutions and funds that promote science and technology in Brazil, such as CNPq, FINEP, P and D ANEEL and Regional Foundations for Research Support. The Bottom-Up methodology consisted in obtaining data directly from the research groups granted by those agencies. After accounting the total Brazilian investment in the period, this was compared with the investments made by the other BRIC countries (Russia, India and China). Next, BRIC countries investment was compared with those made by the European Union, Japan and the United States. The results show that in order to participate in the market share related to equipment and services for the hydrogen economy, Brazil needs to increase the efforts in research, development and innovation in the area. It will be also necessary to apply resources in other important research issues besides ethanol reforming, polymer electrolyte and solid oxide fuel cells, which are the current technologies supported by the Brazilian funding agencies. To achieve this, resources that are already available could be used more efficiently. Another important evidence is that the total annual investment made BRIC countries together is of the same order of magnitude as the investments made separately by the European Union, Japan and the United States. (orig.)

  6. Summary of literature review of risk communication: Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Byram, S.J.

    1991-05-01

    The Hanford Environmental Dose Reconstruction (HEDR) Project will estimate radiation exposures people may have received from radioactive materials released during past operations at the Department of Energy's Hanford Site near Richland, Washington. The project is being conducted by Pacific Northwest Laboratory (PNL) under the direction of an independent Technical Steering Panel (TSP). The Centers for Disease Control (CDC) will use HEDR dose estimates in studies to investigate a potential link between thyroid disease and historical Hanford emissions. The HEDR Project was initiated to address public concerns about the possible health impacts from past releases of radioactive materials from Hanford. The TSP recognized early in the project that special mechanisms would be required to communicate effectively to the many different concerned audiences. To identify and develop these mechanisms, the TSP issued Directive 89-7 to PNL in May 1989. The TSP directed PNL to examine methods to communicate the causes and effects of uncertainties in the dose estimates. A literature review was conducted as the first activity in response to the TSP's directive. This report presents the results of the literature review. The objective of the literature review was to identify ''key principles'' that could be applied to develop communications strategies for the project. 26 refs., 6 figs

  7. Final Project Closeout Report for Sprint Hydrogen Fuel Cell (HFC) Deployment Project in California, Gulf Coast and Eastern Seaboard Markets

    Energy Technology Data Exchange (ETDEWEB)

    Kenny, Kevin [Sprint, Reston, VA (United States); Bradley, Dwayne [Burns & McDonnell, Kansas City, MO (United States)

    2015-09-01

    Sprint is one of the telecommunications industry leaders in the deployment of hydrogen fuel cell (HFC) systems to provide backup power for their mission critical wireless network facilities. With several hundred fuel cells commissioned in California, states in the gulf coast region, and along the upper eastern seaboard. A strong incentive for advancing the integration of fuel cells into the Sprint network came through the award of a Department of Energy (DOE) grant focused on Market Transformation activities for project (EE0000486). This grant was funded by the 2009 American Recovery and Reinvestment Act (ARRA). The funding provided by DOE ($7.295M) was allocated to support the installation of 260 new HFC systems, equipped with an on-site refillable Medium Pressure Hydrogen Storage Solution (MPHSS), as well as for the conversion of 21 low pressure hydrogen systems to the MPHSS, in hopes of reducing barriers to market acceptance.

  8. Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (1). Planning, fuel design, pin fabrication

    International Nuclear Information System (INIS)

    Morihira, Masayuki; Ozawa, Takayuki; Tomita, Yutaka; Suzuki, Masahiro; Kihara, Yoshiyuki; Shigetome, Yoshiaki; Kohno, Shusaku

    2004-07-01

    The collaboration project concerning sphere-pac fuel among JNC, Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) is in progress. Final target of the project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet type fuel and vipack fuel. Total 16 fuel segments (8 pins) of these three types of fuel are planned to be irradiated. Two sphere-pac fuel segments contain 5%Np in addition to 20%Pu-MOX. Other segments contain no Np. The objective of the irradiation tests is to obtain the restructuring data in the early beginning of life for SPF as well as power-to-melt test data for the potential study of SPF. At the same time introduction of modeling technique for irradiation performance analysis, fuel design, fuel fabrication is also important objective for JNC. Fabrication of irradiation test pins was completed till May 2003 in PSI. After transportation of the fuel pins to Petten, two times of irradiation were performed in January to March in 2004 and now post irradiation tests are in progress. Later two irradiations will be done till the autumn in 2004. This report summarized the basic plan, fuel design, and fabrication of irradiation test pins concerning this collaboration project. (author)

  9. OECD/NEA Sandia Fuel Project phase I: Benchmark of the ignition testing

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina, E-mail: martina_adorni@hotmail.it [UNIPI (Italy); Herranz, Luis E. [CIEMAT (Spain); Hollands, Thorsten [GRS (Germany); Ahn, Kwang-II [KAERI (Korea, Republic of); Bals, Christine [GRS (Germany); D' Auria, Francesco [UNIPI (Italy); Horvath, Gabor L. [NUBIKI (Hungary); Jaeckel, Bernd S. [PSI (Switzerland); Kim, Han-Chul; Lee, Jung-Jae [KINS (Korea, Republic of); Ogino, Masao [JNES (Japan); Techy, Zsolt [NUBIKI (Hungary); Velazquez-Lozad, Alexander; Zigh, Abdelghani [USNRC (United States); Rehacek, Radomir [OECD/NEA (France)

    2016-10-15

    Highlights: • A unique PWR spent fuel pool experimental project is analytically investigated. • Predictability of fuel clad ignition in case of a complete loss of coolant in SFPs is assessed. • Computer codes reasonably estimate peak cladding temperature and time of ignition. - Abstract: The OECD/NEA Sandia Fuel Project provided unique thermal-hydraulic experimental data associated with Spent Fuel Pool (SFP) complete drain down. The study conducted at Sandia National Laboratories (SNL) was successfully completed (July 2009 to February 2013). The accident conditions of interest for the SFP were simulated in a full scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate severe accident code validation and to reduce modeling uncertainties within the codes. Phase I focused on axial heating and burn propagation in a single PWR 17 × 17 assembly (i.e. “hot neighbors” configuration). Phase II addressed axial and radial heating and zirconium fire propagation including effects of fuel rod ballooning in a 1 × 4 assembly configuration (i.e. single, hot center assembly and four, “cooler neighbors”). This paper summarizes the comparative analysis regarding the final destructive ignition test of the phase I of the project. The objective of the benchmark is to evaluate and compare the predictive capabilities of computer codes concerning the ignition testing of PWR fuel assemblies. Nine institutions from eight different countries were involved in the benchmark calculations. The time to ignition and the maximum temperature are adequately captured by the calculations. It is believed that the benchmark constitutes an enlargement of the validation range for the codes to the conditions tested, thus enhancing the code applicability to other fuel assembly designs and configurations. The comparison of

  10. Fiscal Year (FY) 2017 Activities for the Spent Fuel Nondestructive Assay Project

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McMath, Garrett Earl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Grogan, Brandon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-11

    The main focus of research in the NA-241 spent fuel nondestructive assay (NDA) project in FY17 has been completing the fabrication and testing of two prototype instruments for upcoming spent fuel measurements at the Clab interim storage facility in Sweden. One is a passive instrument: Differential Die-away Self Interrogation-Passive Neutron Albedo Reactivity (DDSI), and one is an active instrument: Differential Die-Away-Californium Interrogation with Prompt Neutron (DDA). DDSI was fabricated and tested with fresh fuel at Los Alamos National Laboratory in FY15 and FY16, then shipped to Sweden at the beginning of FY17. Research was performed in FY17 to simplify results from the data acquisition system, which is complex because signals from 56 different 3He detectors must be processed using list mode data. The DDA instrument was fabricated at the end of FY16. New high count rate electronics better suited for a spent fuel environment (i.e., KM-200 preamplifiers) were built specifically for this instrument in FY17, and new Tygon tubing to house electrical cables was purchased and installed. Fresh fuel tests using the DDA instrument with numerous configurations of fuel rods containing depleted uranium (DU), low enriched uranium (LEU), and LEU with burnable poisons (Gd) were successfully performed and compared to simulations.1 Additionally, members of the spent fuel NDA project team travelled to Sweden for a “spent fuel characterization and decay heat” workshop involving simulations of spent fuel and analysis of uncertainties in decay heat calculations.

  11. Dose comparison of different scan projections of Implagraphy cone beam computed tomography for dental maxillofacial use

    International Nuclear Information System (INIS)

    Fang Dong; Yuan Xianshun; Zhang Dongsheng

    2012-01-01

    Objective: To evaluate the subject's absorbed dose,equivalent dose and effective dose. Methods: The CBCT unit was Implagraphy and three scan projections were selected such as mandible, maxilla and temporomandibular joint (TMJ). Thermoluminescent dosimeter tubes were used to record the absorbed dose at special positions in the head and neck region of an adult skull and tissue-equivalent phantom. 16 interested organs included pituitary, lens, parotid glands, submandibular glands, sublingual glands, diploe, spongy bone of the chin and cervical vertebra, skins of cheeks and nuchal region, thyroid and esophagus. The absorbed dose was measured in these organs, and then the effective dose (E 1990 , E 2007 ) were calculated according to different ICRP tissue weighting factors. Results: The absorbed dose of mandible,maxilla and TMJ scan varied from (0.99 ±0.09) to (12.85 ±0.09)mGy, (0.93 ±0.01) to (13.07 ±0.02) mGy and (0.68 ±0.01) to (10.18 ± 0.04)mGy. There was significant difference among the three scan projections (F=19.61-30992.27, P<0.05). The equivalent doses of lens and skin were (1.11± 0.07)-(5.76 ± 0.06) mSv and (6.96 ± 0.06)-(10.64 ± 0.07) mSv. There was significant difference among the three scan projections (F=4473.02, 9385.50, P<0.05). The effective dose (E 1990 , E 2007 ) was [(191.35±1.53), (325.17 ±2.58) μSv] for mandible scan, [(106.62 ±2.17), (226.28 ±2.81) μSv] for maxilla scan, [(104.21 ± 1.02), (142.36 ± 1.90) μSv]for TMJ scan, respectively. Conclusions: The valid measurement should be taken to reduce the subject's dose such as a careful history and clinical examination before the performance of CBCT, the latest risk/benefit assessment,precise scan position, the shielding of thyroid as well as brain and the smaller volume size as well. (authors)

  12. Experience with respect to dose limitation in nuclear fuel service operations in the United Kingdom supporting civil nuclear power programmes

    International Nuclear Information System (INIS)

    Kennedy, J.W.

    1983-01-01

    Within the United Kingdom, the nuclear power generation programme is supported by nuclear fuel services including uranium enrichment, fuel fabrication and reprocessing, operated by British Nuclear Fuels Limited (BNFL). These have entailed the processing of large quantities of uranium and of plutonium and fission products arising in the course of irradiation of fuel in nuclear power stations and have necessitated substantial programmes for the radiological protection of the public and of the workers employed in the industry. This paper presents and reviews the statistics of doses recorded in the various sectors of nuclear fuel services operations against the background of the standards to which the industry is required to operate. A description is given of the development of BNFL policy in keeping with the objective of being recognized as among those industries regarded as safe and the resource implications of measures to reduce doses received by workers are reviewed in the light of experience. Finally, the paper reviews the epidemiological data which have been, and continue to be, collected for workers who have been employed in these nuclear fuel services. (author)

  13. International evaluation of Swedish research projects on the environmental impacts of wood fuel harvesting

    Energy Technology Data Exchange (ETDEWEB)

    Hornung, M [Inst. of Terrestrial Ecology, Grange-over-Sands (United Kingdom); Kellomaeki, S [Joensuu Univ. (Finland). Dept. of Forestry; Larsen, J B [Royal Veterinary Univ., Fredriksberg (Denmark). Dept. of Economics and Natural Resources

    1996-12-31

    The purpose of this evaluation was to inform NUTEK of the scientific quality of the research projects, as seen in an international context. The projects were therefore the main elements considered in the evaluation. The main basis of the evaluation was the scientific quality of the research and its relevance to NUTEK`s aims in the application of industrial research and development. The present report is based on the information contained in the written reports submitted by the grant holders, site visits and discussions between the grant holders and the Committee. The report first gives an overview and general recommendations concerning the overall programme on the Environmental Impacts of Wood Fuel Harvest. Thereafter, the projects are evaluated separately. The Committee was unanimous in its conclusions. Evaluated projects: Whole tree harvesting effects on forest soil; Whole tree utilization - forest yield; Nature conservation/Forest energy; Utilizing hardwoods from first thinnings of spruce as fuel wood

  14. International evaluation of Swedish research projects on the environmental impacts of wood fuel harvesting

    Energy Technology Data Exchange (ETDEWEB)

    Hornung, M. [Inst. of Terrestrial Ecology, Grange-over-Sands (United Kingdom); Kellomaeki, S. [Joensuu Univ. (Finland). Dept. of Forestry; Larsen, J.B. [Royal Veterinary Univ., Fredriksberg (Denmark). Dept. of Economics and Natural Resources

    1995-12-31

    The purpose of this evaluation was to inform NUTEK of the scientific quality of the research projects, as seen in an international context. The projects were therefore the main elements considered in the evaluation. The main basis of the evaluation was the scientific quality of the research and its relevance to NUTEK`s aims in the application of industrial research and development. The present report is based on the information contained in the written reports submitted by the grant holders, site visits and discussions between the grant holders and the Committee. The report first gives an overview and general recommendations concerning the overall programme on the Environmental Impacts of Wood Fuel Harvest. Thereafter, the projects are evaluated separately. The Committee was unanimous in its conclusions. Evaluated projects: Whole tree harvesting effects on forest soil; Whole tree utilization - forest yield; Nature conservation/Forest energy; Utilizing hardwoods from first thinnings of spruce as fuel wood

  15. International evaluation of Swedish research projects on the environmental impacts of wood fuel harvesting

    International Nuclear Information System (INIS)

    Hornung, M.; Kellomaeki, S.; Larsen, J.B.

    1995-01-01

    The purpose of this evaluation was to inform NUTEK of the scientific quality of the research projects, as seen in an international context. The projects were therefore the main elements considered in the evaluation. The main basis of the evaluation was the scientific quality of the research and its relevance to NUTEK's aims in the application of industrial research and development. The present report is based on the information contained in the written reports submitted by the grant holders, site visits and discussions between the grant holders and the Committee. The report first gives an overview and general recommendations concerning the overall programme on the Environmental Impacts of Wood Fuel Harvest. Thereafter, the projects are evaluated separately. The Committee was unanimous in its conclusions. Evaluated projects: Whole tree harvesting effects on forest soil; Whole tree utilization - forest yield; Nature conservation/Forest energy; Utilizing hardwoods from first thinnings of spruce as fuel wood

  16. Selection of dominant radionuclides for Phase 1 of the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Napier, B.A.

    1991-07-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions from nuclear operations at Hanford since their inception in 1944. A vital step in the estimation of radiation doses is the determination of the source term,'' that is, the quantities of radionuclides that were released to the environment from the various Hanford operations. Hanford operations have at various times involved hundreds of different radionuclides, some in relatively large quantities. Those radionuclides present in the largest quantities, although significant from an operational handling point of view, may not necessarily have been those of greatest concern for offsite radiation dose. This report documents the selection of the dominant radionuclides (those that may have resulted in the largest portion of the received doses) in the source term for Phase 1 of the HEDR Project, that is, for atmospheric releases from 1944 through 1947 and for surface water releases from 1964 through 1966. 15 refs., 3 figs., 10 tabs.

  17. Overview of fuel testing capabilities at the OECD Halden reactor project

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W [Institutt for Atomenergi, Halden (Norway). OECD Halden Reaktor Projekt

    1994-12-31

    Fuel performance and reliability investigations at the OECD Haiden Reactor Project are described. They are supported by a variety of irradiation rigs, suitable irradiation techniques and a range of instrumentation. Testing capabilities and applications are mainly aimed at exploring mechanisms of fuel behaviour and high burnup. Examples of fuel performance taken from data provided by the Halden Project for the IAEA Co-ordinated Research Programme FUMEX are presented. A number of heavily instrumented rigs to suit different test objects have been developed: base irradiation rig, gas meter rig, diameter measurement rig, ramp rig, gas flow rig, instrumented fuel assembly. In core-measurements and variety of sensors as : fuel thermocouples, bellows pressure transducers, fuel stack elongation detectors, cladding diameter gauge and cladding elongation detectors have been used. Techniques which make it possible to obtain reliable data for all relevant burnups from beginning-of-life to ultra high exposure reaching 100 Mwd/kg UO{sub 2} are described. 7 figs., 3 refs.

  18. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  19. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  20. General guidelines for the assessment of internal dose from monitoring data: Progress of the IDEAS project

    International Nuclear Information System (INIS)

    Doerfel, H.; Andrasi, A.; Bailey, M.; Blanchardon, E.; Cruz-Suarez, R.; Berkovski, V.; Castellani, C. M.; Hurtgenv, C.; Leguen, B.; Malatova, I.; Marsh, J.; Stather, J.; Zeger, J.

    2007-01-01

    In recent major international intercomparison exercises on intake and internal dose assessments from monitoring data, the results calculated by different participants varied significantly. Based on this experience the need for harmonisation of the procedures has been formulated within an EU 5. Framework Programme research project. The aim of the project, IDEAS, is to develop general guidelines for standardising assessments of intakes and internal doses. The IDEAS project started in October 2001 and ended in June 2005. The project is closely related to some goals of the work of Committee 2 of the ICRP and since 2003 there has been close cooperation between the two groups. To ensure that the guidelines are applicable to a wide range of practical situations, the first step was to compile a database of well-documented cases of internal contamination. In parallel, an improved version of an existing software package was developed and distributed to the partners for further use. A large number of cases from the database was evaluated independently by the partners and the results reviewed. Based on these evaluations, guidelines were drafted and discussed with dosimetry professionals from around the world by means of a virtual workshop on the Internet early in 2004. The guidelines have been revised and refined on the basis of the experiences and discussions in this virtual workshop. The general philosophy of the Guidelines is presented here, focusing on the principles of harmonisation, optimisation and proportionality. Finally, the proposed Levels of Task to structure the approach of internal dose evaluation are reported. (authors)

  1. UPS CNG Truck Fleet Start Up Experience: Alternative Fuel Truck Evaluation Project

    International Nuclear Information System (INIS)

    Walkowicz, K.

    2001-01-01

    UPS operates 140 Freightliner Custom Chassis compressed natural gas (CNG)-powered vehicles with Cummins B5.9G engines. Fifteen are participating in the Alternative Fuel Truck Evaluation Project being funded by DOE's Office of Transportation Technologies and the Office of Heavy Vehicle Technologies

  2. International R and D project on development of coated particle fuel for innovative reactors

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    The paper presents an outline for an international collaborative project of coated particle fuel development for innovative reactors. Specific issues include identification of R and D needs and the Member State facilities for meeting the needs followed by development and demonstration of technology. (author)

  3. 76 FR 70955 - Helena Nation Forest: Dalton Mountain Forest Restoration & Fuels Reduction Project

    Science.gov (United States)

    2011-11-16

    ... DEPARTMENT OF AGRICULTURE Forest Service Helena Nation Forest: Dalton Mountain Forest Restoration & Fuels Reduction Project AGENCY: Forest Service, USDA. ACTION: Notice of intent to prepare an environmental impact statement. SUMMARY: The Helena National Forest (HNF) is proposing on the Lincoln Ranger...

  4. Description of Website for the OECD-IAEA Paks Fuel Project

    International Nuclear Information System (INIS)

    Szabo, Emese; Hozer, Zoltan; Nagy, Imre

    2010-01-01

    The first version of the database for the OECD-IAEA PAKS FUEL PROJECT has been collected and it is available on the following password protected website: http://nagy.aeki.kfki.hu Several modifications have been made and new items added according to the minutes of the 1st meeting held in Budapest on 30-31 January 2005

  5. Integrating fuel treatment into ecosystem management: A proposed project planning process

    Science.gov (United States)

    Keith D. Stockmann; Kevin D. Hyde; J. Greg Jones; Dan R. Loeffler; Robin P. Silverstein

    2010-01-01

    Concern over increased wildland fire threats on public lands throughout the western United States makes fuel reduction activities the primary driver of many management projects. This single-issue focus recalls a management planning process practiced frequently in recent decades - a least-harm approach where the primary objective is first addressed and then plans are...

  6. Experiences from Swedish demonstration projects with phosphoric acid fuel cells; Erfarenheter fraan svenska demonstrationsprojekt med fosforsyrabraensleceller

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Per [Sycon Energikonsult AB, Stockholm (Sweden); Sarkoezi, Laszlo [Vattenfall Utveckling AB, Stockholm (Sweden)

    1999-10-01

    In Sweden, there are today two phosphoric acid fuel cells installed, one PC25A which have been in operation in more than 4 years, and one PC25C which have been in operation for two years. The aim with this project has been two compare operation characteristics, performance, and operation experiences for these two models.

  7. Projection of fossil fuels consumption in the Venezuelan electricity generation industry

    International Nuclear Information System (INIS)

    Vidoza, Jorge A.; Gallo, Waldyr L.R.

    2016-01-01

    This study presents a prospective analysis on the impacts of recent efficient energy policies application in Venezuela, integrating both oil production and electricity supply to assess energy resources balance in a quantitative manner. A special focus is given to main fossil fuels used in the electric power industry; natural gas, diesel oil and fuel oil. Four scenarios were proposed, ranging from a low-economy-growth/low-efficiency scenario to an optimist high-economy-growth/high-efficiency scenario. Efficiency effects are more notorious for high-economy-growth case, fuel consumption for electricity generation reduces 38% for natural gas, 12% for diesel and 29% for fuel oil, in the established time period. Deficits in oil and gas Venezuelan production were also determined, deficits are highly affected by economical forecasting, and by fuel smuggling in Venezuelan borders. Results showed the high importance of energy efficiency policies development for Venezuela, in order to reduce fossil fuel domestic consumption to allocate them in a more profitable market. - Highlights: • We made a prospective analysis on efficient energy policies impacts in Venezuela. • Reduced fuel consumption was obtained for efficient scenarios. • Current energy regulations are not enough to encourage energy efficiency. • Hydroelectricity projects need more promotion to have deeper impacts.

  8. Rescue dose orders as an alternative to range orders: an evidence-based practice project.

    Science.gov (United States)

    Yi, Cassia

    2015-06-01

    Relief of pain is a fundamental aspect of optimal patient care. However, pain management in the inpatient setting is often constrained by concerns related to regulatory oversight, particularly with regard to the use of opioid dose range orders. These concerns can inadvertently result in the development of policies and practices that can negatively impact the health care team's ability to deliver optimal and individualized pain management. An evidence-based practice project was undertaken to address concerns about regulatory oversight of pain management processes by changing the way pain was managed in a large academic hospital setting. A novel pain management approach using rescue dose medications was established as an alternative to opioid dose range orders. The use of the rescue dose protocol was successfully implemented. Outcomes included an overall reduction in the administration of inappropriate intravenous opioids and opioid-acetaminophen combination medications, with a subsequent increase in single-entity first-line opioid analgesics. Rescue dose protocols may offer an alternative to opioid dose range orders as a means of effectively managing pain. Copyright © 2015 American Society of PeriAnesthesia Nurses. Published by Elsevier Inc. All rights reserved.

  9. Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

    International Nuclear Information System (INIS)

    Horvath, M.; Munoz-Reja, C.; Tverberg, T.; Jenssen, H. K.

    2010-01-01

    Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP). HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD - Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as a partner in the fuel research programs. Improving the NPP operations, fuel cycles were designed to increase fuel burnup. Higher fuel burnup reduces the number of spent fuel assemblies and thus the costs of new fuel as well as the costs of back-end management. Higher burnup is reached either by prolonging the reactor cycles or by increasing the number of reactor cycles for the fuel in the core. Both ways entail additional requirements concerning fuel enrichment and burnable absorbers as additives and adjustments on the cladding material properties, such as mechanical treatment and chemical composition of the alloys. For these demands and needs ENUSA promotes the research on high burnup effects, gadolinium doped fuels and cladding material behaviour under irradiation. Various experiments, called IFA, are developed and performed also by providing materials. ENUSA collaborates with HRP on various experiments investigating the fuel densification and swelling, fission gas release, pressure limits on UO 2 and (U,Gd)O 2 fuels (IFA-504, -515, -636, -681); the cladding creep, lift-off, corrosion and hydrides on different tubing materials (IFA-567, -610, -638); instrumentation of the experiments, especially on pre-irradiated materials (IFA-533). These experiments are combined with model calculations to improve predictions for higher burnups and to maintain safety margins (IFA-515, -636, -681). Besides these unique in-pile experiments PIEs are performed as well on fuel and structural materials to complete the scope of these studies (IFA

  10. A spatially encoded dose difference maximal intensity projection map for patient dose evaluation: A new first line patient quality assurance tool

    Energy Technology Data Exchange (ETDEWEB)

    Hu Weigang; Graff, Pierre; Boettger, Thomas; Pouliot, Jean [Department of Radiation Oncology, University of California, San Francisco, San Francisco, California 94143 (United States); and others

    2011-04-15

    Purpose: To develop a spatially encoded dose difference maximal intensity projection (DD-MIP) as an online patient dose evaluation tool for visualizing the dose differences between the planning dose and dose on the treatment day. Methods: Megavoltage cone-beam CT (MVCBCT) images acquired on the treatment day are used for generating the dose difference index. Each index is represented by different colors for underdose, acceptable, and overdose regions. A maximal intensity projection (MIP) algorithm is developed to compress all the information of an arbitrary 3D dose difference index into a 2D DD-MIP image. In such an algorithm, a distance transformation is generated based on the planning CT. Then, two new volumes representing the overdose and underdose regions of the dose difference index are encoded with the distance transformation map. The distance-encoded indices of each volume are normalized using the skin distance obtained on the planning CT. After that, two MIPs are generated based on the underdose and overdose volumes with green-to-blue and green-to-red lookup tables, respectively. Finally, the two MIPs are merged with an appropriate transparency level and rendered in planning CT images. Results: The spatially encoded DD-MIP was implemented in a dose-guided radiotherapy prototype and tested on 33 MVCBCT images from six patients. The user can easily establish the threshold for the overdose and underdose. A 3% difference between the treatment and planning dose was used as the threshold in the study; hence, the DD-MIP shows red or blue color for the dose difference >3% or {<=}3%, respectively. With such a method, the overdose and underdose regions can be visualized and distinguished without being overshadowed by superficial dose differences. Conclusions: A DD-MIP algorithm was developed that compresses information from 3D into a single or two orthogonal projections while hinting the user whether the dose difference is on the skin surface or deeper.

  11. A spatially encoded dose difference maximal intensity projection map for patient dose evaluation: a new first line patient quality assurance tool.

    Science.gov (United States)

    Hu, Weigang; Graff, Pierre; Boettger, Thomas; Pouliot, Jean

    2011-04-01

    To develop a spatially encoded dose difference maximal intensity projection (DD-MIP) as an online patient dose evaluation tool for visualizing the dose differences between the planning dose and dose on the treatment day. Megavoltage cone-beam CT (MVCBCT) images acquired on the treatment day are used for generating the dose difference index. Each index is represented by different colors for underdose, acceptable, and overdose regions. A maximal intensity projection (MIP) algorithm is developed to compress all the information of an arbitrary 3D dose difference index into a 2D DD-MIP image. In such an algorithm, a distance transformation is generated based on the planning CT. Then, two new volumes representing the overdose and underdose regions of the dose difference index are encoded with the distance transformation map. The distance-encoded indices of each volume are normalized using the skin distance obtained on the planning CT. After that, two MIPs are generated based on the underdose and overdose volumes with green-to-blue and green-to-red lookup tables, respectively. Finally, the two MIPs are merged with an appropriate transparency level and rendered in planning CT images. The spatially encoded DD-MIP was implemented in a dose-guided radiotherapy prototype and tested on 33 MVCBCT images from six patients. The user can easily establish the threshold for the overdose and underdose. A 3% difference between the treatment and planning dose was used as the threshold in the study; hence, the DD-MIP shows red or blue color for the dose difference > 3% or < or = 3%, respectively. With such a method, the overdose and underdose regions can be visualized and distinguished without being overshadowed by superficial dose differences. A DD-MIP algorithm was developed that compresses information from 3D into a single or two orthogonal projections while hinting the user whether the dose difference is on the skin surface or deeper.

  12. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    International Nuclear Information System (INIS)

    Wiesenack, W.

    1996-01-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project's data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup

  13. Initial communication survey results for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Beck, D.M.

    1991-03-01

    To support the public communication efforts of the Technical Steering Panel of the Hanford Environmental Dose Reconstruction (HEDR) Project, a public survey was conducted. The survey was intended to provide information about the public's knowledge and interest in the project and the best ways to communicate project results. Questions about the project were included as part of an omnibus survey conducted by Washington State University. The survey was conducted by phone to Washington State residents in the spring of 1990. This report gives the HEDR-related questions and summary data of responses. Questions associated with the HEDR Project were grouped into four categories: knowledge of the HEDR Project; interest in the project; preferred ways of receiving information about the project (including public information meetings, a newsletter mailed to homes, presentations to civic groups in the respondent's community, a computer bulletin board respondent could access with a modem, information displays at public buildings and shopping malls, and an information video sent to respondent); and level of concern over past exposure from Hanford operations. Questions abut whom state residents are most likely to trust about radiation issues were also part of the omnibus survey, and responses are included in this report

  14. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

  15. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2003-02-12

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

  16. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    PICKETT, W.W.

    2000-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure

  17. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2001-05-15

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted in section 3

  18. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2003-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

  19. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2001-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted in section 3.1.5 and will be

  20. Some radioactivity concentrations and ingestion dose projections arising from consumption of food containing Chernobyl contamination

    International Nuclear Information System (INIS)

    Paz, L.R. de la; Palattao, M.V.; Estacio, J.F.L.; Anden, A.

    1987-04-01

    Doses arising from the ingestion of radioactive contamination coming from Chernobyl accident are calculated using various radioactivity limits adopted by different organizations after the accident. These are compared with that allowed in the Philippines. Projected concentrations of Cs-137 and Cs-134 in various food items in the affected countries, one month and one year after the accident are calculated using a model proposed by Boone, Ng and Palms. Except for food produced in one or two hot spots, the projected concentrations after one year are expected to return to within the range of pre-Chernobyl values. (Auth.) 12 refs.; 13 tabs.; 6 figs

  1. A real-time stack radioactivity monitoring system and dose projection program

    Energy Technology Data Exchange (ETDEWEB)

    Hull, A.P.; Michael, P.A. [Brookhaven National Laboratory, Upton, NY (United States); Bernstein, H.J. [Bernstein & Sons, Bellport, NY (United States)

    1995-02-01

    At Brookhaven National Laboratory, a commercial Low- and High-Range Air Effluent Monitor has become operational at the 60 Mw (t) High Flux Beam Reactor. Its output data is combined with that from ground-level and elevated meteorological sensors to provide a real-time projection of the down-wind dose rates from noble gases and radioiodines released from the HFBR`s 100 m stack. The output of the monitor, and the meteorological sensors and the dose projections can be viewed at emergency response terminals located in the Reactor Control Room, its Technical Support Center and at the laboratory`s separately located Meteorological Station and Monitoring and Assessment Center.

  2. Computational model design specification for Phase 1 of the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Napier, B.A.

    1991-07-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emission from nuclear operations at Hanford since their inception in 1944. The purpose of this report is to outline the basic algorithm and necessary computer calculations to be used to calculate radiation doses specific and hypothetical individuals in the vicinity of Hanford. The system design requirements, those things that must be accomplished, are defined. The system design specifications, the techniques by which those requirements are met, are outlined. Included are the basic equations, logic diagrams, and preliminary definition of the nature of each input distribution. 4 refs., 10 figs., 9 tabs

  3. Computational model design specification for Phase 1 of the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Napier, B.A.

    1991-07-01

    The objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emission from nuclear operations at Hanford since their inception in 1944. The purpose of this report is to outline the basic algorithm and necessary computer calculations to be used to calculate radiation doses specific and hypothetical individuals in the vicinity of Hanford. The system design requirements, those things that must be accomplished, are defined. The system design specifications, the techniques by which those requirements are met, are outlined. Included are the basic equations, logic diagrams, and preliminary definition of the nature of each input distribution. 4 refs., 10 figs., 9 tabs.

  4. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    International Nuclear Information System (INIS)

    B. McLeod

    2002-01-01

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M and O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M and O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report

  5. Development of Graphical User Interface for ARRBOD (Acute Radiation Risk and BRYNTRN Organ Dose Projection)

    Science.gov (United States)

    Kim, Myung-Hee; Hu, Shaowen; Nounu, Hatem N.; Cucinotta, Francis A.

    2010-01-01

    The space radiation environment, particularly solar particle events (SPEs), poses the risk of acute radiation sickness (ARS) to humans; and organ doses from SPE exposure may reach critical levels during extra vehicular activities (EVAs) or within lightly shielded spacecraft. NASA has developed an organ dose projection model using the BRYNTRN with SUMDOSE computer codes, and a probabilistic model of Acute Radiation Risk (ARR). The codes BRYNTRN and SUMDOSE, written in FORTRAN, are a Baryon transport code and an output data processing code, respectively. The ARR code is written in C. The risk projection models of organ doses and ARR take the output from BRYNTRN as an input to their calculations. BRYNTRN code operation requires extensive input preparation. With a graphical user interface (GUI) to handle input and output for BRYNTRN, the response models can be connected easily and correctly to BRYNTRN in friendly way. A GUI for the Acute Radiation Risk and BRYNTRN Organ Dose (ARRBOD) projection code provides seamless integration of input and output manipulations, which are required for operations of the ARRBOD modules: BRYNTRN, SUMDOSE, and the ARR probabilistic response model. The ARRBOD GUI is intended for mission planners, radiation shield designers, space operations in the mission operations directorate (MOD), and space biophysics researchers. The ARRBOD GUI will serve as a proof-of-concept example for future integration of other human space applications risk projection models. The current version of the ARRBOD GUI is a new self-contained product and will have follow-on versions, as options are added: 1) human geometries of MAX/FAX in addition to CAM/CAF; 2) shielding distributions for spacecraft, Mars surface and atmosphere; 3) various space environmental and biophysical models; and 4) other response models to be connected to the BRYNTRN. The major components of the overall system, the subsystem interconnections, and external interfaces are described in this

  6. Breast compression and radiation dose in two different mammographic oblique projections: 45 and 60 deg

    International Nuclear Information System (INIS)

    Brnic, Zoran; Hebrang, Andrija

    2001-01-01

    Introduction: Standard mammography includes two views, craniocaudal and medio-lateral oblique. Depending on patient's body constitution, central beam angle in mediolateral oblique projection may vary, with 45 deg. being suitable for the majority of patients in routine daily practice. With continuous improvement in X-ray technology and radiographers' training, the risk of radiation induced cancerogenesis is considerably reduced and acceptable when compared to benefit. However, the risk still exists, being cumulative and directly related to absorbed glandular dose. There is no minimal dose of radiation which is absolutely harmless, and every effort to reduce the dose is welcome. In this retrospective study two different angles (45 vs. 60 deg.) of mediolateral oblique view were compared according to radiation dose and efficacy of breast compression. Patients and methods: In 52 women, additional 60 deg. oblique films were done after craniocaudal and mediolateral oblique 45 deg.-films, with the same kVp and positioning technique. Breast thickness, time-current products (mA s) and absorbed doses were compared between 45 deg. - and 60 deg.-films. Subgroups of women with large, small, prominent and pendulous breasts were analyzed separately, following the same methodology as for the whole group. Results: mA s were 11.5% lower and compression 7% better with an angle of 60 deg. than with 45 deg. In the subgroup of women with small breasts, mA s values were 13% lower and compression 9% better with 60 deg. than with 45 deg., while in the subgroup with large breasts, mA s were 9% lower and compression 5% better. In the subgroup of patients with pendulous breasts, mA s values were 12% lower and compression 10% better with 60 deg. than with 45 deg., while in the subgroup with prominent breasts, mA s values were 4% lower and compression 3% better. Absorbed glandular dose was estimated to be approximately 20% lower when an oblique mammogram was done with 60 deg. instead of 45 deg

  7. Dose estimation and prediction of radiation effects on aquatic biota resulting from radioactive releases from the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Witherspoon, J.P.

    1975-01-01

    Aquatic organisms are exposed to radionuclides released to the environment during various steps of the nuclear fuel cycle. Routine releases from these processes are limited in compliance with technical specifications and requirements of federal regulations. These regulations reflect I.C.R.P. recommendations which are designed to provide an environment considered safe for man. It is generally accepted that aquatic organisms will not receive damaging external radiation doses in such environments; however, because of possible bioaccumulation of radionuclides there is concern that aquatic organisms might be adversely affected by internal doses. The objectives of this paper are: to estimate the radiation dose received by aquatic biota from the different processes and determine the major dose-contributing radionuclides, and to assess the impact of estimated doses on aquatic biota. Dose estimates are made by using radionuclide concentration measured in the liquid effluents of representative facilities. This evaluation indicates the potential for the greatest radiation dose to aquatic biota from the nuclear fuel supply facilities (i.e., uranium mining and milling). The effects of chronic low-level radiation on aquatic organisms are discussed from somatic and genetic viewpoints. Based on the body of radiobiological evidence accumulated up to the present time, no significant deleterious effects are predicted for populations of aquatic organisms exposed to the estimated dose rates resulting from routine releases from conversion, enrichment, fabrication, reactors and reprocessing facilities. At the doses estimated for milling and mining operations it would be difficult to detect radiation effects on aquatic populations; however, the significance of such radiation exposures to aquatic populations cannot be fully evaluated without further research on effects of chronic low-level radiation. (U.S.)

  8. Environmental Assessment for Lignite Fuel Enhancement Project, Coal Creek Station, Great River Energy, Underwood, North Dakota

    Energy Technology Data Exchange (ETDEWEB)

    N/A

    2004-01-16

    The US Department of Energy (DOE) has prepared this EA to assess the environmental impacts of the commercial application of lignite fuel enhancement. The proposed demonstration project would be implemented at Great River Energy's Coal Creek Station near Underwood, North Dakota. The proposed project would demonstrate a technology to increase the heating value of lignite and other high-moisture coals by reducing the moisture in the fuels. Waste heat that would normally be sent to the cooling towers would be used to drive off a percentage of the moisture contained within the lignite. Application of this technology would be expected to boost power-generating efficiencies, provide economic cost savings for lignite and sub-bituminous power plants, and reduce air emissions. The proposed project would be constructed on a previously disturbed site within the Coal Creek Station and no negative impacts would occur in any environmental resource area.

  9. Data base on dose reduction research projects for nuclear power plants

    International Nuclear Information System (INIS)

    Khan, T.A.; Yu, C.K.; Roecklein, A.K.

    1994-05-01

    This is the fifth volume in a series of reports that provide information on dose reduction research and health physics technology or nuclear power plants. The information is taken from two of several databases maintained by Brookhaven National Laboratory's ALARA Center for the Nuclear Regulatory Commission. The research section of the report covers dose reduction projects that are in the experimental or developmental phase. It includes topics such as steam generator degradation, decontamination, robotics, improvements in reactor materials, and inspection techniques. The section on health physics technology discusses dose reduction efforts that are in place or in the process of being implemented at nuclear power plants. A total of 105 new or updated projects are described. All project abstracts from this report are available to nuclear industry professionals with access to a fax machine through the ACEFAX system or a computer with a modem and the proper communications software through the ACE system. Detailed descriptions of how to access all the databases electronically are in the appendices of the report

  10. Data base on dose reduction research projects for nuclear power plants. Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Khan, T.A.; Yu, C.K.; Roecklein, A.K. [Brookhaven National Lab., Upton, NY (United States)

    1994-05-01

    This is the fifth volume in a series of reports that provide information on dose reduction research and health physics technology or nuclear power plants. The information is taken from two of several databases maintained by Brookhaven National Laboratory`s ALARA Center for the Nuclear Regulatory Commission. The research section of the report covers dose reduction projects that are in the experimental or developmental phase. It includes topics such as steam generator degradation, decontamination, robotics, improvements in reactor materials, and inspection techniques. The section on health physics technology discusses dose reduction efforts that are in place or in the process of being implemented at nuclear power plants. A total of 105 new or updated projects are described. All project abstracts from this report are available to nuclear industry professionals with access to a fax machine through the ACEFAX system or a computer with a modem and the proper communications software through the ACE system. Detailed descriptions of how to access all the databases electronically are in the appendices of the report.

  11. A preliminary examination of audience-related communications issues for the Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, C.W.

    1991-04-01

    The Hanford Environmental Dose Reconstruction (HEDR) Project will estimate radiation doses people may have received from exposure to radioactive materials released during past operations at the US Department of Energy's (DOE) Hanford Site near Richland, Washington. The HEDR Project was initiated in response to public concerns about possible health impacts from past releases of radioactive materials from Hanford. The TSP recognized early in the project that special mechanisms would be required to effectively communicate to the many different concerned audiences. Accordingly, the TSP directed PNL to examine methods for communicating causes and effects of uncertainties in the dose estimates. After considering the directive and discussing it with the Communications Subcommittee of the TSP, PNL undertook a broad investigation of communications methods to consider for inclusion in the TSP's current communications program. As part of this investigation, a literature review was conducted regarding risk communications. A key finding was that, in order to successfully communicate risk-related information, a thorough understanding of the knowledge level, concerns and information needs of the intended recipients (i.e., the audience) is necessary. Hence, a preliminary audience analysis was conducted as part of the present research. This report summarizes the results of this analysis. 1 ref., 9 tabs.

  12. Disaggregation of collective dose-a worked example based on future discharges from the Sellafield nuclear fuel reprocessing site, UK

    International Nuclear Information System (INIS)

    Jones, S R; Lambers, B; Stevens, A

    2004-01-01

    Collective dose has long been advocated as an important measure of the detriment associated with practices that involve the use of radioactivity. Application of collective dose in the context of worker protection is relatively straightforward, whereas its application in the context of discharges to the environment can yield radically different conclusions depending upon the population groups and integration times that are considered. The computer program PC-CREAM98 has been used to provide an indicative disaggregation into individual dose bands of the collective dose due to potential future radioactive discharges from the nuclear fuel reprocessing site at Sellafield in the UK. Two alternative discharge scenarios are considered, which represent a 'stop reprocessing early, minimum discharge' scenario and a 'reprocessing beyond current contracts' scenario. For aerial discharges, collective dose at individual effective dose rates exceeding 0.015 μSv y -1 is only incurred within the UK, and at effective dose rates exceeding 1.5 μSv y -1 is only incurred within about 20 km of Sellafield. The geographical distribution of collective dose from liquid discharges is harder to assess, but it appears that collective dose incurred outside the UK is at levels of individual effective dose rate below 1.5 μSv y -1 , with the majority being incurred at rates of 0.002 μSv y -1 or less. In multi-attribute utility analyses, the view taken on the radiological detriment to be attributed to the two discharge scenarios will depend critically on the weight or monetary value ascribed to collective doses incurred within the differing bands of individual dose rate

  13. Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J.; Cooper, T.D.

    1995-10-20

    The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ``stabilized`` form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision.

  14. Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1

    International Nuclear Information System (INIS)

    Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J.; Cooper, T.D.

    1995-01-01

    The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ''stabilized'' form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision

  15. General guidelines for the Assessment of Internal Dose from Monitoring Data (Project IDEAS)

    International Nuclear Information System (INIS)

    Doerfel, H.; Andrasi, A.; Bailey, M.; Blanchardon, E.; Berkovski, V.; Castellani, C. M.; Hurtgen, C.; Jourdain, J. R.; LeGuen, B.; Puncher, M.

    2004-01-01

    In recent major international intercomparison exercises on intake and internal dose assessments from monitoring data the results calculated by different participants varied significantly. This was mainly due to the broad variety of methods and assumptions applied in the assessment procedure. Based on these experiences the need for harmonisation of the procedures has been formulated within an EU research project under the 5th Framework Programme. The aim of the project, IDEAS, is to develop general guidelines for standardising assessments of intakes and internal doses. The IDEAS project started in October 2001 and will end in March 2005. Eight institutions from seven European countries are participating. Inputs from internal dosimetry professionals from across Europe are also being used to ensure a broad consensus in the outcome of the project. The IDEAS project is closely related to some goals of the work of Committee 2 of the ICRP and since 2003 there has been close cooperation between the two groups. To ensure that the guidelines are applicable to a wide range of practical situations, the first step has been to compile a database of well-documented cases of internal contamination. In parallel, an improved version of an existing software package has been developed and distributed to the partners for further use. A large number of cases from the database have been evaluated independently by partners in the project using the same software and the results have been reviewed. Based on these evaluations guidelines are being drafted and will be discussed with dosimetry professionals from around the world by means of a virtual workshop on the Internet early in 2004. The guidelines will be revised and refined on the basis of the experiences and discussions of this virtual workshop and the outcome of an intercomparison exercise organised as part of the project. This will be open to all internal dosimetry professionals. (Author) 10 refs

  16. Integrated Data Base for 1989: Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1989-11-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1988. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, commercial reactor and fuel cycle facility decommissioning waste, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous, highly radioactive materials that may require geologic disposal. 45 figs., 119 tabs

  17. Integrated data base for 1990: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1990-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1989. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 22 refs., 48 figs., 109 tabs

  18. Integrated Data Base for 1991: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1991-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1990. These data are based on the most reliable information available form government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated generally through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 160 refs., 61 figs., 142 tabs

  19. Commercial milk distribution profiles and production locations. Hanford Environmental Dose Reconstruction Project

    Energy Technology Data Exchange (ETDEWEB)

    Deonigi, D.E.; Anderson, D.M.; Wilfert, G.L.

    1994-04-01

    The Hanford Environmental Dose Reconstruction (HEDR) Project was established to estimate radiation doses that people could have received from nuclear operations at the Hanford Site since 1944. For this period iodine-131 is the most important offsite contributor to radiation doses from Hanford operations. Consumption of milk from cows that ate vegetation contaminated by iodine-131 is the dominant radiation pathway for individuals who drank milk (Napier 1992). Information has been developed on commercial milk cow locations and commercial milk distribution during 1945 and 1951. The year 1945 was selected because during 1945 the largest amount of iodine-131 was released from Hanford facilities in a calendar year (Heeb 1993); therefore, 1945 was the year in which an individual was likely to have received the highest dose. The year 1951 was selected to provide data for comparing the changes that occurred in commercial milk flows (i.e., sources, processing locations, and market areas) between World War II and the post-war period. To estimate the doses people could have received from this milk flow, it is necessary to estimate the amount of milk people consumed, the source of the milk, the specific feeding regime used for milk cows, and the amount of iodine-131 contamination deposited on feed.

  20. Radiation shielding and dose rate evaluation at the interim storage facility for spent fuel from Cernavoda NPP

    International Nuclear Information System (INIS)

    Stanciu, Marcela; Mateescu, Silvia; Pantazi, Doina; Penescu, Maria

    2000-01-01

    At present studies necessary to license the Interim Storage Facility for the Spent Fuel (CANDU type) from Cernavoda NPP are developed in our country.The spent fuel from Cernavoda NPP is discharged into Spent Fuel Bay in Service Building of the plant, where it remains several years for cooling. After this period, the bundles of spent fuel are to be transferred to the Interim Storage Facility.The dry interim storage solution seems to be the most appropriate variant for Cernavoda NPP.The design of the Spent Fuel Interim Storage Facility must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility achievement. In this paper we intend to present the calculation of radiation shielding at the spent fuel interim storage facility for two technical solutions: - Concrete Monolithic Module and Concrete Storage Cask. In order to quantify the fuel composition after irradiation, the isotope generation and depletion code ORIGEN 2.1 has been used, taking into account a cooling time of 7 years and 9 years, respectively, for these two variants. The shielding calculations have been performed using the computer codes QAD-5K and MICROSHIELD-4. The evaluations refer only to gamma radiation because the resulting neutron source (from (α,n) reactions and spontaneous fission) is insignificant as compared to the gamma source. The final results consist in the minimum thickness of the shielding and the corresponding external dose rates, ensuring a design average dose rate based on national and international regulations. (authors)

  1. FY 1993 task plans for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Shipler, D.B.

    1991-10-01

    The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses from Hanford Site operations since 1944 to individuals and populations. The primary objective of work to be performed in FY 1993 is to complete the source term estimates and dose estimates for key radionuclides for the air and river pathways. At the end of FY 1993, the capability will be in place to estimate doses for individuals in the extended (32-county) study area, 1944--1991. Native American research will continue to provide input for tribal dose estimates. In FY 1993, the Technical Steering Panel (TSP) will decide whether demographic and river pathways data collection should be extended beyond FY 1993 levels. The FY 1993 work scopes and milestones in this document are based on the work plan discussed at the TSP Budget/Fiscal Subcommittee meeting on August 19--20, 1991. Table 1 shows the FY 1993 milestones; Table 2 shows estimated costs. The subsequent work scope descriptions are based on the milestones. This document and the FY 1992 task plans will form the basis for a contract with Battelle and the Centers for Disease Control (CDC). The 2-year dose reconstruction contract is expected to begin in February 1992. This contract will replace the current arrangement, whereby the US Department of Energy directly funds the Pacific Northwest Laboratory to conduct dose reconstruction work. In late FY 1992, the FY 1993 task plans will be more fully developed with detailed technical approaches, data quality objectives, and budgeted labor hours. The task plans will be updated again in July 1993 to reflect any scope, milestone, or cost changes directed during the year by the TSP. 2 tabs

  2. Spent fuel performance data: An analysis of data relevant to the NNWSI Project

    International Nuclear Information System (INIS)

    Oversby, V.M.; Shaw, H.F.

    1987-08-01

    This paper summarizes the physical and chemical properties of spent light water reactor fuel that might influence its performance as a waste form under geologic disposal conditions at Yucca Mountain, Nevada. Results obtained on the dissolution testing of spent fuel conducted by the NNWSI Project are presented and discussed. Work published by other programs, in particular those of Canada and Sweden, are reviewed and compared with the NNWSI testing results. An attempt is made to relate all of the results to a common basis of presentation and to rationalize apparent conflicts between sets of results obtained under different experimental conditions

  3. Uncertainty in projected climate change arising from uncertain fossil-fuel emission factors

    Science.gov (United States)

    Quilcaille, Y.; Gasser, T.; Ciais, P.; Lecocq, F.; Janssens-Maenhout, G.; Mohr, S.

    2018-04-01

    Emission inventories are widely used by the climate community, but their uncertainties are rarely accounted for. In this study, we evaluate the uncertainty in projected climate change induced by uncertainties in fossil-fuel emissions, accounting for non-CO2 species co-emitted with the combustion of fossil-fuels and their use in industrial processes. Using consistent historical reconstructions and three contrasted future projections of fossil-fuel extraction from Mohr et al we calculate CO2 emissions and their uncertainties stemming from estimates of fuel carbon content, net calorific value and oxidation fraction. Our historical reconstructions of fossil-fuel CO2 emissions are consistent with other inventories in terms of average and range. The uncertainties sum up to a ±15% relative uncertainty in cumulative CO2 emissions by 2300. Uncertainties in the emissions of non-CO2 species associated with the use of fossil fuels are estimated using co-emission ratios varying with time. Using these inputs, we use the compact Earth system model OSCAR v2.2 and a Monte Carlo setup, in order to attribute the uncertainty in projected global surface temperature change (ΔT) to three sources of uncertainty, namely on the Earth system’s response, on fossil-fuel CO2 emission and on non-CO2 co-emissions. Under the three future fuel extraction scenarios, we simulate the median ΔT to be 1.9, 2.7 or 4.0 °C in 2300, with an associated 90% confidence interval of about 65%, 52% and 42%. We show that virtually all of the total uncertainty is attributable to the uncertainty in the future Earth system’s response to the anthropogenic perturbation. We conclude that the uncertainty in emission estimates can be neglected for global temperature projections in the face of the large uncertainty in the Earth system response to the forcing of emissions. We show that this result does not hold for all variables of the climate system, such as the atmospheric partial pressure of CO2 and the

  4. Ultra-Clean Fischer-Tropsch Fuels Production and Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Steve Bergin

    2005-10-14

    The Report Abstract provides summaries of the past year's activities relating to each of the main project objectives. Some of the objectives will be expanded on in greater detail further down in the report. The following objectives have their own addition sections in the report: Dynamometer Durability Testing, the Denali Bus Fleet Demonstration, Bus Fleet Demonstrations Emissions Analysis, Impact of SFP Fuel on Engine Performance, Emissions Analysis, Feasibility Study of SFPs for Rural Alaska, and Cold Weather Testing of Ultra Clean Fuel.

  5. Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

    International Nuclear Information System (INIS)

    Newsom, H.C.

    1999-01-01

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted

  6. Spent nuclear fuel project multi-year work plan WBS number 1.4.1

    International Nuclear Information System (INIS)

    Wells, J.L.

    1997-01-01

    The Spent Nuclear Fuel (SNF) Project Multi-Year Work Plan (MYWP) is a controlled living document that contains the current SNF Project Technical, Schedule and Cost Baselines. These baselines reflect the current Project execution strategies and are controlled via the change control process. Other changes to the MYWP document will be controlled using the document control process. These changes will be processed as they are approved to keep the MYWP a living document. The MYWP will be maintained continuously as the project baseline through the life of the project and not revised annually. The MYWP is the one document which summarizes and links these three baselines in one place. Supporting documentation for each baseline referred to herein may be impacted by changes to the MYWP, and must also be revised through change control to maintain consistency

  7. Landscape modeling for dose calculations in the safety assessment of a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Lindborg, Tobias; Kautsky, Ulrik; Brydsten, Lars

    2007-01-01

    The Swedish Nuclear Fuel and Waste Management Co.,(SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site. Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the

  8. Calculation of Water Levels in Spent Fuel Pool and Effective Dose Followed by the Worker Geometrically Exposed to Radiation using Gamma-ray Source

    International Nuclear Information System (INIS)

    Lee, Donghee; Park, Kwangheon; Yoon, Hyoungju

    2013-01-01

    If the total effective dose value is lower than the surface dose rate of the water, the worker is able to work in a safe environment. In the case that the level of spent fuel pool is up to 550cm, there exists the limitation for workers to access to the storage pool because the result value is about 8 times higher than surface dose rate. In the case that the level of spent fuel pool is higher than 600cm, however, it can be safe work environment because the result value is lower than surface dose rate. Therefore, in the case of ISO geometry which is the same with practical situation, when considering Gamma-ray emission from spent fuel, effective dose is much higher than surface dose rate when the level of storage pool is lower than the height of fuel, 452.8cm. On the other hand, the level of effective dose decreases rapidly when the level of storage pool is higher than the level of the fuel. This means that it is not the safe environment when the level of fuel below 140cm is lower than surface dose rate. That is why the access of workers should be limited. Whereas, in the case of the level of storage pool above 600cm which is about 140cm higher than the level of the fuel, it is the safe environment for workers because the result value becomes lower than surface dose rate As a result, the level of wet storage of spent fuel should be at least 600cm for workers to work in safe environment because lower dose than surface dose rate makes less radiation exposure

  9. Measurement of dose rates and Monte Carlo analysis of neutrons in a spent-fuel shipping vessel

    International Nuclear Information System (INIS)

    Ueki, K.; Namito, Y.; Fuse, T.

    1986-01-01

    On-board experiments were carried out in a spent-fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox 3 were loaded in five holds, and neutron and gamma-ray dose rates were measured on the hatch covers of the holds. Before shipping those casks, dose rates were also measured on the cask surfaces, one by one, to eliminate radiation from other casks. The Monte Carlo coupling technique was employed successfully to analyze the measured neutron dose rate distributions in the spent-fuel shipping vessel. Through this study, the Monte Carlo coupling code system, MORSE-CG/CASK-VESSEL, on which the MORSE-CG code was based, was established. The agreement between the measured and the calculated neutron dose rates on the TN-12A cask surface was quite satisfactory. The calculated neutron dose rates agreed with the measured values within a factor of 1.5 on the hold 3 hatch cover and within a factor of 2 on the hold 5 hatch cover in which the concrete shield was fixed in the Pacific Swan

  10. FY 1992 task plans for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    1991-10-01

    Phase 1 of the HEDR Project was designed to develop and demonstrate a method for estimating radiation doses people may have received from Hanford Site operations since 1944. The method researchers developed relied on a variety of measured and reconstructed data as input to a modular computer model that generates dose estimates and their uncertainties. As part of Phase 1, researchers used the reconstructed data and computer model to calculate preliminary dose estimates for populations from limited radionuclides, in a limited geographical area and time period. Phase 1 ended in FY 1990. In February 1991, the TSP decided to shift the project planning approach away from phases--which were centered around completion of major portions of technical activities--to individual fiscal years (FYs), which span October of one year through September of the next. Therefore, activities that were previously designated to occur in phases are now designated in an integrated schedule to occur in one or more of the next fiscal years into FY 1995. Task plans are updated every 6 months. In FY 1992, scientists will continue to improve Phase 1 data and models to calculate more accurate and precise dose estimates. The plan for FY 1992 has been prepared based on activities and budgets approved by the Technical Steering Panel (TSP) at its meeting on August 19--20, 1991. The activities can be divided into four categories: (1) model and data evaluation activities, (2) additional dose estimates, (3) model and data development activities, and (4) technical and communication support. 3 figs., 2 tabs

  11. Transportation Energy Futures Series. Projected Biomass Utilization for Fuels and Power in a Mature Market

    Energy Technology Data Exchange (ETDEWEB)

    Ruth, M. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Mai, T. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Newes, E. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Aden, A. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Warner, E. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Uriarte, C. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Inman, D. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Simpkins, T. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Argo, A. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2013-03-01

    The viability of biomass as transportation fuel depends upon the allocation of limited resources for fuel, power, and products. By focusing on mature markets, this report identifies how biomass is projected to be most economically used in the long term and the implications for greenhouse gas (GHG) emissions and petroleum use. In order to better understand competition for biomass between these markets and the potential for biofuel as a market-scale alternative to petroleum-based fuels, this report presents results of a micro-economic analysis conducted using the Biomass Allocation and Supply Equilibrium (BASE) modeling tool. The findings indicate that biofuels can outcompete biopower for feedstocks in mature markets if research and development targets are met. The BASE tool was developed for this project to analyze the impact of multiple biomass demand areas on mature energy markets. The model includes domestic supply curves for lignocellulosic biomass resources, corn for ethanol and butanol production, soybeans for biodiesel, and algae for diesel. This is one of a series of reports produced as a result of the Transportation Energy Futures (TEF) project, a Department of Energy-sponsored multi-agency project initiated to pinpoint underexplored strategies for abating GHGs and reducing petroleum dependence related to transportation.

  12. Transportation Energy Futures Series: Projected Biomass Utilization for Fuels and Power in a Mature Market

    Energy Technology Data Exchange (ETDEWEB)

    Ruth, M.; Mai, T.; Newes, E.; Aden, A.; Warner, E.; Uriarte, C.; Inman, D.; Simpkins, T.; Argo, A.

    2013-03-01

    The viability of biomass as transportation fuel depends upon the allocation of limited resources for fuel, power, and products. By focusing on mature markets, this report identifies how biomass is projected to be most economically used in the long term and the implications for greenhouse gas (GHG) emissions and petroleum use. In order to better understand competition for biomass between these markets and the potential for biofuel as a market-scale alternative to petroleum-based fuels, this report presents results of a micro-economic analysis conducted using the Biomass Allocation and Supply Equilibrium (BASE) modeling tool. The findings indicate that biofuels can outcompete biopower for feedstocks in mature markets if research and development targets are met. The BASE tool was developed for this project to analyze the impact of multiple biomass demand areas on mature energy markets. The model includes domestic supply curves for lignocellulosic biomass resources, corn for ethanol and butanol production, soybeans for biodiesel, and algae for diesel. This is one of a series of reports produced as a result of the Transportation Energy Futures (TEF) project, a Department of Energy-sponsored multi-agency project initiated to pinpoint underexplored strategies for abating GHGs and reducing petroleum dependence related to transportation.

  13. Spent fuel performance assessment and research. Final report of a co-ordinated research project on Spent Fuel Performance Assessment and Research (SPAR) 1997-2001

    International Nuclear Information System (INIS)

    2003-03-01

    The report provides an overview of technical issues related to spent fuel wet and dry storage and summarizes the objectives and major findings of research, carried out within the framework of the Coordinated Research Program. Included are the fuel integrity aspects, fuel degradation mechanisms in dry and wet storage, behaviour of storage facility components (metallic components, reinforced concrete). Also included are issues related to long-term storage and monitoring technologies and techniques. Country reports on research projects within the SPAR Coordinated Research Program is presented. A brief history is given on the history of the BEFAST and SPAR Coordinated Research Projects

  14. Spent fuel performance assessment and research. Final report of a co-ordinated research project on Spent Fuel Performance Assessment and Research (SPAR) 1997-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    The report provides an overview of technical issues related to spent fuel wet and dry storage and summarizes the objectives and major findings of research, carried out within the framework of the Coordinated Research Program. Included are the fuel integrity aspects, fuel degradation mechanisms in dry and wet storage, behaviour of storage facility components (metallic components, reinforced concrete). Also included are issues related to long-term storage and monitoring technologies and techniques. Country reports on research projects within the SPAR Coordinated Research Program is presented. A brief history is given on the history of the BEFAST and SPAR Coordinated Research Projects.

  15. Approaches in estimation of external cost for fuel cycles in the ExternE project

    International Nuclear Information System (INIS)

    Afanas'ev, A.A.; Maksimenko, B.N.

    1998-01-01

    The purposes, content and main results of studies realized within the frameworks of the International Project ExternE which is the first comprehensive attempt to develop general approach to estimation of external cost for different fuel cycles based on utilization of nuclear and fossil fuels, as well as on renewable power sources are discussed. The external cost of a fuel cycle is treated as social and environmental expenditures which are not taken into account by energy producers and consumers, i.e. these are expenditures not included into commercial cost nowadays. The conclusion on applicability of the approach suggested for estimation of population health hazards and environmental impacts connected with electric power generation growth (expressed in money or some other form) is made

  16. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    Directory of Open Access Journals (Sweden)

    Bourg Stéphane

    2015-12-01

    Full Text Available Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes. Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.

  17. Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

    International Nuclear Information System (INIS)

    Horvath, M. I.; Jenssen, H. K.; Munoz-Reja, C.; Tverberg, T.

    2011-01-01

    Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP), HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD-Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as partner in the fuel research programs. Various experiments are developed and performed also by providing materials, ENUSA collaborates with HRP on various experiments investigating the fuel behaviour, especially on Gd-bearing fuel. 20 years of successful collaboration between HRP and ENUSA is continuing with promising and results to ensure and enhance the safe operation of the Spanish and all other NPPs in the world. (Author) 12 refs.

  18. A multiscale filter for noise reduction of low-dose cone beam projections.

    Science.gov (United States)

    Yao, Weiguang; Farr, Jonathan B

    2015-08-21

    The Poisson or compound Poisson process governs the randomness of photon fluence in cone beam computed tomography (CBCT) imaging systems. The probability density function depends on the mean (noiseless) of the fluence at a certain detector. This dependence indicates the natural requirement of multiscale filters to smooth noise while preserving structures of the imaged object on the low-dose cone beam projection. In this work, we used a Gaussian filter, exp(-x2/2σ(2)(f)) as the multiscale filter to de-noise the low-dose cone beam projections. We analytically obtained the expression of σ(f), which represents the scale of the filter, by minimizing local noise-to-signal ratio. We analytically derived the variance of residual noise from the Poisson or compound Poisson processes after Gaussian filtering. From the derived analytical form of the variance of residual noise, optimal σ(2)(f)) is proved to be proportional to the noiseless fluence and modulated by local structure strength expressed as the linear fitting error of the structure. A strategy was used to obtain the reliable linear fitting error: smoothing the projection along the longitudinal direction to calculate the linear fitting error along the lateral direction and vice versa. The performance of our multiscale filter was examined on low-dose cone beam projections of a Catphan phantom and a head-and-neck patient. After performing the filter on the Catphan phantom projections scanned with pulse time 4 ms, the number of visible line pairs was similar to that scanned with 16 ms, and the contrast-to-noise ratio of the inserts was higher than that scanned with 16 ms about 64% in average. For the simulated head-and-neck patient projections with pulse time 4 ms, the visibility of soft tissue structures in the patient was comparable to that scanned with 20 ms. The image processing took less than 0.5 s per projection with 1024   ×   768 pixels.

  19. Spent Nuclear Fuel Project FY 1996 Multi-Year Program Plan WBS No. 1.4.1, Revision 1

    International Nuclear Information System (INIS)

    1995-09-01

    This document describes the Spent Nuclear Fuel (SNF) Project portion of the Hanford Strategic Plan for the Hanford Reservation in Richland, Washington. The SNF Project was established to evaluate and integrate the urgent risks associated with N-reactor fuel currently stored at the Hanford site in the K Basins, and to manage the transfer and disposition of other spent nuclear fuels currently stored on the Hanford site. An evaluation of alternatives for the expedited removal of spent fuels from the K Basin area was performed. Based on this study, a Recommended Path Forward for the K Basins was developed and proposed to the U.S. DOE

  20. Spent Nuclear Fuel Project FY 1996 Multi-Year Program Plan WBS No. 1.4.1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This document describes the Spent Nuclear Fuel (SNF) Project portion of the Hanford Strategic Plan for the Hanford Reservation in Richland, Washington. The SNF Project was established to evaluate and integrate the urgent risks associated with N-reactor fuel currently stored at the Hanford site in the K Basins, and to manage the transfer and disposition of other spent nuclear fuels currently stored on the Hanford site. An evaluation of alternatives for the expedited removal of spent fuels from the K Basin area was performed. Based on this study, a Recommended Path Forward for the K Basins was developed and proposed to the U.S. DOE.

  1. Experimental verification of methods for gamma dose rate calculations in the vicinity of containers with the RA reactor spent fuel elements

    International Nuclear Information System (INIS)

    Milosevic, M.; Cupac, S.; Pesic, M.

    2005-01-01

    The methodology for equivalent gamma dose rate determination on the outer surface of existing containers with the spent fuel elements of the RA reactor is briefly summarised, and experimental verification of this methodology in the field of gamma rays near the aluminium channel with spent fuel elements lifted from the stainless steel containers no. 275 in the RA reactor hall is presented. The proposed methodology is founded on: the existing fuel burnup data base; methods and models for the photon source determination in the RA reactor spent fuel elements developed in the Vinca Institute, and validated Monte Carlo codes for the equivalent gamma dose rate calculations. (author) [sr

  2. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions

  3. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  4. Spent fuel and radioactive waste inventories and projections as of December 31, 1980

    International Nuclear Information System (INIS)

    1981-09-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on the most reliable information available from Government sources and the open literature, technical reports, and direct contacts. Future waste generation rates and characteristics of these materials to be accumulated over the remainder of this century are also presented, based on a present DOE/EIA projection of US commercial nuclear power growth and expected defense-related and industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level wastes, transuranic waste, low-level waste, remedial action waste, active uranium mill tailings, and airborne waste. For each category, current and projected inventories are given through the year 2000. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for the present inventories of inactive uranium mill tailings

  5. Spent fuel and radioactive waste: an integrated data base of inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Notz, K.J.; Forsberg, C.W.; Mastal, E.F.

    1984-01-01

    The Integrated Data Base (IDB) Program provides official US Department of Energy (DOE) data on spent fuel and radioactive waste inventories, projections, and characteristics. This information is provided through the cooperative efforts of the IDB Program and DOE lead offices, lead sites, major programs, and generator sites. The program is entering its fifth year, and major accomplishments are summarized in three broad areas: (1) the annual inventory report, including ORIGEN2 applications and a Quality Assurance (QA) plan; (2) the summary data file and direct user access; and (3) data processing methodology and support to other programs. Plans for future work in these areas are outlined briefly, including increased utilization of personal computers. Some examples of spent fuel data are given in terms of projected quantities for two growth scenarios, burnup and age profile of the existing inventory, and the approximate specific thermal power relative to high-level waste (HLW) from various sources. 4 refs., 2 figs., 3 tabs

  6. Atmospheric transport and dispersion modeling for the Hanford Environmental Dose Reconstruction Project

    International Nuclear Information System (INIS)

    Ramsdell, J.V.

    1991-07-01

    Radiation doses that may have resulted from operations at the Hanford Site are being estimated in the Hanford Environmental Dose Reconstruction (HEDR) Project. One of the project subtasks, atmospheric transport, is responsible for estimating the transport, diffusion and deposition of radionuclides released to the atmosphere. This report discusses modeling transport and diffusion in the atmospheric pathway. It is divided into three major sections. The first section of the report presents the atmospheric modeling approach selected following discussion with the Technical Steering Panel that directs the HEDR Project. In addition, the section discusses the selection of the MESOI/MESORAD suite of atmospheric dispersion models that form the basis for initial calculations and future model development. The second section of the report describes alternative modeling approaches that were considered. Emphasis is placed on the family of plume and puff models that are based on Gaussian solution to the diffusion equations. The final portion of the section describes the performance of various models. The third section of the report discusses factors that bear on the selection of an atmospheric transport modeling approach for HEDR. These factors, which include the physical setting of the Hanford Site and the available meteorological data, serve as constraints on model selection. Five appendices are included in the report. 39 refs., 4 figs., 2 tabs

  7. Preliminary design report: Prototypical Spent Fuel Consolidation Equipment Demonstration Project: Phase 1

    International Nuclear Information System (INIS)

    Blissell, W.H.; Ciez, A.P.; Mitchell, J.L.; Winkler, C.J.

    1986-12-01

    This document describes the Westinghouse Preliminary Design for the Prototypical Consolidation Demonstration Project per Department of Energy (DOE) Contract No. DE-AC07-86ID12649 and under direction of the DOE Idaho Operations Office. The preliminary design is the first step to providing the Department of Energy with a fully qualified, licensable, cost-effective spent fuel rod consolidation system. The design was developed using proven technologies and equipment to create an innovative approach to previous rod consolidation concepts. These innovations will better enable the Westinghouse system to: consolidate fuel rods in a precise, fully-controlled, accountable manner; package all rods from two PWR fuel assemblies or from four BWR fuel assemblies in one 8.5 inch square consolidated rods canister; meet all functional requirements; operate with all fuel types common to the US commercial nuclear industry with minimal tooling changeouts; and meet consolidation production process rates, while maintaining operator and public health and safety. This Preliminary Design Report provides both detailed descriptions of the equipment required to perform the rod consolidation process and the supporting analyses to validate the design

  8. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1979-12-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed hear are of a continuing nature. Results and conclusions described may change as the work continues

  9. Fuel cells in railway systems. Project report; Brennstoffzellen im Schienenverkehr. Projektbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-06-01

    The project report describes the state of the art of fuel cell technology and the specific requirements of railway systems. Potential applications in the state of Schleswig-Holstein are gone into. [German] In diesem Projektbericht wird der Stand der Brennstoffzellentechnik und die Entwicklung von Anforderungen des Schienenverkehrs an diese Technik ermittelt. Darueber hinaus wird untersucht, inwieweit eine Modellregion Schleswig-Holstein zur weiteren Entwicklung der Technik beitragen kann.

  10. CARA Project: development of the advanced ULE fuel element for heavy water nuclear power plants

    International Nuclear Information System (INIS)

    Brasnarof, Daniel O.; Marino, Armando C.; Florido, Pablo C.; Munoz, C.; Bianchi, Daniel R.; Giorgis, Miguel A.

    2006-01-01

    The CARA Project (Spanish acronym of Combustible Avanzado para Reactores Argentinos) is a national fuel element technology development, compatible with our nuclear power plants (Atucha I, Embalse and Atucha II). It takes into account the experience obtained in our nuclear organisations (CNEA-CONUAR-NASA). The goal of the CARA fuel element is the performance improvement for those reactors and the enhancing of their normal operative conditions. The CARA design allows the burnup extension by using 52 rods of the same diameter. Likewise it keeps good thermo-hydraulic behaviour. The fuel bundle can be directly used in nuclear power plants with horizontal channels. By using an additional system it can be installed in the PHWR with vertical channels. The expected profits, by the use of the CARA in our reactors, broadly guaranty the recovery of the fund for its development, due to a reduction of the NPP fuels and back end cost. We estimate a reduction in the generation cost between 20 or 25 % in relation to the present one if we use 0.85 or 0.90% SEU (Slightly Enriched Uranium). The use of the CARA fuel in our reactors will also reduce the amount of spent fuel to be treated. The shortening could be between 17 to 27 % in Atucha I in relation to the present ULE (0.85%), between 38 to 46% for Embalse, and 45 to 53% for Atucha II. The mechanical behaviour and hydraulic compatibility have been verified. Several CARA prototypes were fabricated with a new design of the end plate and with new processes for the welding for the rods. We present in this paper the current status of the CARA fuel element development. (author) [es

  11. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1996-12-01

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens

  12. Tribal and stakeholder communication and participation strategy for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    Hoofer, V.L.

    1995-12-01

    This document outlines a plan to ensure the effective involvement of the Hanford stakeholders and Tribal Governments in Spent Nuclear Fuel (SNF) Project issues and decisions. Stakeholders are defined as the public, news media, regulators, employees, Hanford Advisory Board and members of local, state, and federal governments. Experience at Hanford has shown that early and continued involvement of all interested parties in decision making is absolutely essential for fostering project success. Failure to recognize the importance of this interaction has resulted in significant cost in terms of time and money for several site programs

  13. Externalities of fuel cycles 'ExternE' project. Natural gas fuel cycle. Estimation of physical impacts and monetary valuation for priority impact pathways

    International Nuclear Information System (INIS)

    Holland, M.; Watkiss, P.; Berry, J.; Johnson, C.; Lee, D.

    1994-01-01

    This document assesses the progress made in quantifying environmental and health damages associated with the natural gas fuel cycle for electricity generation. The methodology developed in the ExternE Project is described in more detail elsewhere (European Commission, 1994a; 1995, in preparation). The reader is referred to these earlier reports for wider discussion of many of the issues underlying this type of work. The increased desire for economic assessment of environmental damage reflects growing awareness of problems such as global warming, ozone depletion and the acidification and nutrification of ecosystems. A wide range of receptors are affected, including human health, forests, crops, and buildings. Such damages are typically not accounted for by the producers and consumers of the good in question (in this case energy). They are thus referred to as 'external costs' or 'externalities', to distinguish them from the private costs which account for the construction of plant, cost of fuel, wages, etc. At the political level there are a variety of reasons for the growing interest in the quantification of the environmental impacts of energy use and the related external costs. These include the need to integrate environmental concerns when selecting between different fuels and energy technologies and the need to evaluate the costs and benefits of stricter environmental standards. These issues are reflected in European Union policy, through, for example, the Maastricht Treaty, the 5th Environmental Action Programme 'towards sustainability', the European Commission's White Paper 'Growth, competitiveness, employment and ways forward to the 21st century' and the establishment of the European Environmental Agency. The proposal for an Energy-Carbon tax is the first concrete proposal by the European Union for the direct use of economic instruments in environmental policy in the energy sector. An agreed methodology for calculation and integration of external costs has not

  14. Gamma dose rate calculations for conceptual design of a shield system for spent fuel interim dry storage in CNA 1

    International Nuclear Information System (INIS)

    Blanco, A; Gomez S

    2012-01-01

    After completing the rearrangement of the Spent Fuel Elements (SFE) into a compact arrangement in the two storage water pools, Atucha Nuclear Reactor 1 (ANR 1) will leave free position for the wet storage of the SFE discharged until December 2014. Even, in two possible scenarios, such as extending operation from 2015 or the cessation of operation after that date, it will be necessary to empty the interim storage water pools transferring the SFE to a temporary dry storage system. Because the law 25.018 'Management of Radioactive Wastes' implies for the first scenario - operation beyond 2015 - that Nucleoelectrica Argentina S.A. will still be in charge of the dry storage system and for the second - the cessation of operation after 2015 - the National Commission of Atomic Energy (CNEA) will be in charge by the National Management Program of Radioactive Wastes, the interim dry storage system of SNF is an issue of common interest which justifies go forward together. For that purpose and in accordance with the criticality and shielding calculations relevant to the project, in this paper we present the dose rate calculations for shielding conceptual design of a system for dry interim storage of the SFE of ANR 1. The specifications includes that the designed system must be suitable without modification for the SFE of the ANR 2. The results for the calculation of the photon dose rate, in touch and at one meter far, for the Transport Module and the Container of the SFE, are presented, which are required and controlled by the National Regulatory Authority (NRA) and the International Atomic Energy Agency (IAEA). It was used the SAS4 module of SCALE5.1 system and MCNP5. As a design tool for the photon shielding in order to meet current standards for allowable dose rates, a radial and axial parametric analysis were developed based on the thickness of lead of the Transport Module. The results were compared and verified between the two computing systems. Before this

  15. Integrated data base for 1988: Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1988-09-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1987. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reportd for miscellaneous, highly radioactive materials that may require geologic disposal. 89 refs., 46 figs., 104 tabs

  16. Integrated data base for 1986: spent fuel and radioactive waste inventories, projections, and characteristics. Revision 2

    International Nuclear Information System (INIS)

    1986-09-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US Department of Energy (DOE) radioactive wastes through December 31, 1985, based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. Current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the expected defense-related and private industrial and institutional activities and the latest DOE/Energy Information Administration (EIA) projections of US commercial nuclear power growth. The materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or calculated isotopic compositions

  17. Spent Nuclear Fuel Project document control and Records Management Program Description

    International Nuclear Information System (INIS)

    MARTIN, B.M.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project document control and records management program, as defined within this document, is based on a broad spectrum of regulatory requirements, Department of Energy (DOE) and Project Hanford and SNF Project-specific direction and guidance. The SNF Project Execution Plan, HNF-3552, requires the control of documents and management of records under the auspices of configuration control, conduct of operations, training, quality assurance, work control, records management, data management, engineering and design control, operational readiness review, and project management and turnover. Implementation of the controls, systems, and processes necessary to ensure compliance with applicable requirements is facilitated through plans, directives, and procedures within the Project Hanford Management System (PHMS) and the SNF Project internal technical and administrative procedures systems. The documents cited within this document are those which directly establish or define the SNF Project document control and records management program. There are many peripheral documents that establish requirements and provide direction pertinent to managing specific types of documents that, for the sake of brevity and clarity, are not cited within this document

  18. Calculation of the external dose rate in the spent fuel pool for the case to use compact racks

    International Nuclear Information System (INIS)

    Passos, E.M. dos; Alves, A.S.M.

    1988-01-01

    The possible introduction of compact racks in the spent fuel pool of the Angra 1 Nuclear Power Plant largely inreases its storage capacity, but originates an increase of the gamma radiation sources. The precise evaluation of the effects of the adoption of this option on the external gamma dose rates and also on the thickness of the concrete shielding requires the utilization of sofisticated computer codes (QAD, ANISN), which allow the calculation of the gamma dose rates through thick shielding walls. This paper describes the utilized methodology for the calculation of the modified pool shieldings, showing the obtained results for the Angra 1 NPP case. The gamma dose rate was calculated with the point Kernel model, first analytically, and later through utilization of the tridimensional multigroup QAD computer code. (author) [pt

  19. New generation of nuclear fuels: Stability of different stearates under high doses gamma irradiation in the manufacturing process

    Energy Technology Data Exchange (ETDEWEB)

    Lebeau, D.; Esnouf, S. [Den-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Gracia, J. [Den-Service d' Etudes des Combustibles et Matériaux à base d' Actinides (SECA), CEA, F-30207 Bagnols-sur-Cèze Cedex (France); Audubert, F. [Den-Service d' Analyse et de Caractérisation du Comportement des Combustibles (SA3C), CEA, F- 13115 Saint-Paul-lez-Durance (France); Ferry, M., E-mail: muriel.ferry@cea.fr [Den-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2017-07-15

    In the future reactors, the pellets radioactivity will increase due to the modification of the plutonium concentration. The stability of the organic additive used as lubricating/deagglomerating agent has thus to be evaluated. Up to now, zinc stearate is employed, but new additives are tested in this study and compared to zinc stearate. In a first part of this paper, the order of magnitude of the dose deposited in the stearates has been estimated. Afterward, three different stearates have been irradiated, using gamma-rays at doses as high as 2000 kGy. Two atmospheres of irradiation were tested, i.e. inert atmosphere and air. Samples were characterized using the following analytical tools: mass spectrometry, thermogravimetry and infrared spectroscopy. The objective is the evaluation of the ageing of these materials. In the nuclear fuel pellets manufacturing context, the candidate which could replace zinc stearate, if this one is too degraded to fulfill its role of lubricant in the pellets of the future manufacturing, has been determined. - Highlights: •Dose deposition estimation for different stearates. •Stearates radiolysis and radio-oxidization at high doses using gamma-rays. •H{sub 2} emission estimation as a function of atmosphere and dose. •Chemical modifications in stearates as a function of atmosphere and dose. •Comparison of three stearates.

  20. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

  1. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    International Nuclear Information System (INIS)

    BAZINET, G.D.

    2000-01-01

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those requirements are noted

  2. Low dose radiation effects: an integrative european approach (Risc-Rad Project) coordinated by the Cea

    International Nuclear Information System (INIS)

    Sabatier, L.

    2006-01-01

    RISC-RAD (Radiosensitivity of Individuals and Susceptibility to Cancer induced by ionizing Radiations) is an Integrated Project funded by the European Commission under 6. Framework Programme / EURATOM. RISC-RAD started on 1. January 2004 for a duration of four years. Coordinated by Cea (Dr Laure Sabatier), it involves 11 European countries (Austria, Denmark, Finland, France, Germany, Ireland, Italy, the Netherlands, Spain, Sweden and the United Kingdom) and 29 research institutions. Objectives: Exposures to low and protracted doses of ionizing radiation are very frequent in normal living environment, at work places, in industry and in medicine. Effects of these exposures on human health cannot be reliably assessed by epidemiological methods, nor is thoroughly understood by biologists. RISC-RAD project proposes to help bridging the gap of scientific knowledge about these effects. To achieve this goal, a necessary key step is to understand the basic mechanisms by which radiation induces cancer. Studying this multistage process in an integrated way, the project offers a new biological approach characterised by and clear-cut and objective-driven scientific policy: the project is focused on the effects of low doses (less than 100 mSv) and protracted doses of radiation. It aims at identifying new parameters that take into account the differences in radiation responses between individuals. A group of modelers works closely with the experimental teams in order to better quantify the risks associated with low and protracted doses. Research work is divided into five work packages interacting closely with each other. WP1 is dedicated to DNA damage. Ionizing Radiation (IR) produce a broad spectrum of base modifications and DNA strand breaks of different kinds, among which double-strand breaks and 'clustered damage' which is thought to be a major feature in biological effectiveness of IR. The aim of Work Package 1 is to improve understanding of the initial DNA damage induced by

  3. Implementation of Agile project management in spent nuclear fuel characterization process

    International Nuclear Information System (INIS)

    Vinas Pena, P.

    2015-01-01

    Full text of publication follows. Spent nuclear fuel characterization (SNFC) is a complex process that covers different areas of analysis and whose final goal is to provide an accurate description of spent nuclear fuel (SNF) status for its future classification for storage and transport. The need to reduce the SNFC processing time maintaining the quality of the product has motivated ENUSA to research and implement Agile project management and human performance techniques. The Agile management techniques are focused in accommodate changes or new requirements in the project during the elaboration process without suffering delays or lose of quality. For its SNF projects ENUSA uses 2 complementary techniques: SCRUM and Kanban. SCRUM methodology is based on divide the process into activities blocks. Each block is a finished part of the final product which allows periodical deliveries of the product and the easy introduction of changes if they are necessary. The characterization process is formed by blocks of activities based on different analysis for every fuel assembly as the existence of leaking rods; the analysis of the structural integrity considering the existence of missing rods, broken or missing grids or grid straps or grid springs...; the corrosion phenomenon on the rod that could affect its integrity during the storage and transport; the burnup of the fuel assembly; the analysis of the rod internal pressure and its effect on rod failure mechanism as creep or on the material embrittlement due to the radial hydride precipitation; the compatibility with the container to avoid operational problems during cask loading and unloading, and any new input based on the regulatory evolution and the industry state of the art. The different analysis can be developed at the same time as they are independent. Kanban methodology consists in a visual representation of the evolution of the process. In a chart, the different activities needed to perform any of the analysis

  4. Data base on dose reduction research projects for nuclear power plants. Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    Khan, T.A.; Vulin, D.S.; Liang, H.; Baum, J.W. [Brookhaven National Lab., Upton, NY (United States)

    1992-08-01

    This is the fourth volume in a series of reports that provide information on dose reduction research and health physics technology for nuclear power plants. The information is taken from a data base maintained by Brookhaven National Laboratory`s ALARA Center for the Nuclear Regulatory Commission. This report presents information on 118 new or updated projects, covering a wide range of activities. Projects including steam generator degradation, decontamination, robotics, improvement in reactor materials, and inspection techniques, among others, are described in the research section of the report. The section on health physics technology includes some simple and very cost-effective projects to reduce radiation exposures. Included in this volume is a detailed description of how to access the BNL data bases which store this information. All project abstracts from this report, as well as many other useful documents, can be accessed, with permission, through our on-line system, ACE. A computer equipped with a modem, or a fax machine is all that is required to connect to ACE. Many features of ACE, including software, hardware, and communications specifics, are explained in this report.

  5. Data base on dose reduction research projects for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Khan, T.A.; Vulin, D.S.; Liang, H.; Baum, J.W. (Brookhaven National Lab., Upton, NY (United States))

    1992-08-01

    This is the fourth volume in a series of reports that provide information on dose reduction research and health physics technology for nuclear power plants. The information is taken from a data base maintained by Brookhaven National Laboratory's ALARA Center for the Nuclear Regulatory Commission. This report presents information on 118 new or updated projects, covering a wide range of activities. Projects including steam generator degradation, decontamination, robotics, improvement in reactor materials, and inspection techniques, among others, are described in the research section of the report. The section on health physics technology includes some simple and very cost-effective projects to reduce radiation exposures. Included in this volume is a detailed description of how to access the BNL data bases which store this information. All project abstracts from this report, as well as many other useful documents, can be accessed, with permission, through our on-line system, ACE. A computer equipped with a modem, or a fax machine is all that is required to connect to ACE. Many features of ACE, including software, hardware, and communications specifics, are explained in this report.

  6. Spent nuclear fuel project multi-canister overpack, additional NRC requirements

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1998-01-01

    The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002

  7. Task 27 -- Alaskan low-rank coal-water fuel demonstration project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    Development of coal-water-fuel (CWF) technology has to-date been predicated on the use of high-rank bituminous coal only, and until now the high inherent moisture content of low-rank coal has precluded its use for CWF production. The unique feature of the Alaskan project is the integration of hot-water-drying (HWD) into CWF technology as a beneficiation process. Hot-water-drying is an EERC developed technology unavailable to the competition that allows the range of CWF feedstock to be extended to low-rank coals. The primary objective of the Alaskan Project, is to promote interest in the CWF marketplace by demonstrating the commercial viability of low-rank coal-water-fuel (LRCWF). While commercialization plans cannot be finalized until the implementation and results of the Alaskan LRCWF Project are known and evaluated, this report has been prepared to specifically address issues concerning business objectives for the project, and outline a market development plan for meeting those objectives.

  8. Annual report of the working group 'fuel pin and fuel element mechanics' of the Institut fuer Reaktortechnik (IRT) of the Technische Hochschule Darmstadt for the Fast Breeder Project

    International Nuclear Information System (INIS)

    Fabian, H.; Humbach, W.; Lassmann, K.; Mueller, J.J.; Preusser, T.; Schmelz, K.

    1978-09-01

    This report comprises six single lectures given at an information meeting organized by the Institut fuer Reaktortechnik der Technischen Hochschule Darmstadt (IRT) in Darmstadt on April 24, 1978. The lectures are an account of work performed at IRT on the mechanics of fuel pins and fuel elements and supported by the Fast Breeder Project (PSB) of KfK. These activities can be broken down into studies of the integral fuel pin (URANUS computer code) and into multidimensional studies of the fuel pin using the finite-element method (FINEL and ZIDRIG computer codes). Moreover, a report is presented of the status of the test facility for simulation of out-of-pile cladding tube loads and of the IRT project on the simulation and analysis of radiation damage. (orig./GL) [de

  9. International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). 2008 progress report

    International Nuclear Information System (INIS)

    2009-02-01

    The purpose of the work is to review the progress of the IAEA international project for innovative reactors and fuel cycle technologies (INPRO). The publication reports about the recognition of INPRO and on general Information on INPRO, its strengths, memberships, collaboration with other international initiatives, the INPRO organization and management and the history of INPRO. The section on the progress of INPRO in 2008 contains task 1: INPRO Methodology, task 2: Assessment Studies, task 3: Nuclear Energy Visions for the 21st Century, task 4: Infrastructure and Institutional Innovation, task 5: Common User Considerations and task 6: Collaborative Projects. Conclusions and New Trends are followed by a bibliography. Annex I deals with the INPRO project management in 2008 and Annex II provides a selection of photographs from 2008. Finally a list of acronyms is provided

  10. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Klein, J.A.; Storch, S.N.; Ashline, R.C.

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal

  11. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1992-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal

  12. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

    2013-01-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  13. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  14. DOD Residential Proton Exchange Membrane (PEM) Fuel Cell Demonstration Program. Volume 2. Summary of Fiscal Year 2001-2003 Projects

    Science.gov (United States)

    2005-09-01

    produced many of the Beatles 1970s recordings. This facility was selected to host the UK PEM demonstration project from a selection of four potential sites...funded the Department of Defense (DOD) Residential PEM Demonstration Project to demonstrate domestically-produced, residential Proton Exchange Membrane...PEM) fuel cells at DOD Facilities. The objectives were to: (1) assess PEM fuel cells’ role in supporting sustainability at military installations

  15. Hanford spent nuclear fuel project recommended path forward, volume III: Alternatives and path forward evaluation supporting documentation

    International Nuclear Information System (INIS)

    Fulton, J.C.

    1994-10-01

    Volume I of the Hanford Spent Nuclear Fuel Project - Recommended Path Forward constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. Volume II provided a comparative evaluation of four Alternatives for the Path Forward and an evaluation for the Recommended Path Forward. Although Volume II contained extensive appendices, six supporting documents have been compiled in Volume III to provide additional background for Volume II

  16. Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

    International Nuclear Information System (INIS)

    Filbert, Wolfgang; Herold, Philipp

    2015-01-01

    The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.

  17. Patient dose assessment in various Interventional radiology and cardiology procedures in Algeria (IAEA regional project results)

    International Nuclear Information System (INIS)

    Khelassi-Toutaoui, Nadia; Merad, Ahmed; Toutaoui, A.E.K.; Bairi, Souad

    2008-01-01

    Full text: Purpose: To evaluate patient doses in Interventional Radiology (IR) and Cardiology (IC) procedures in Algeria, within the framework of an International Atomic Energy Agency (IAEA) regional project on radiation protection of patients and medical exposure control (RAF 9033). Materials and Methods: Three public hospitals (CHU Bab el Oued, CHU Parnet and CHU Mustapha) and one specialised Cardiology Service (Clinique Maouche) were chosen for the study. For Maximum Skin Dose (MSD) evaluation, gafchromic films XR type R were used, placed on patient's back before the procedure. The Dose Area Product (DAP) and MSD were measured in 57 IR and IC procedures, either diagnostic or therapeutic. Results: The results revealed large variations in MSD (0.06-3.3 Gy) and DAP (5.5-332 mGycm 2 ). Mean MSD was 0.227 Gy in cerebral angiography, 0.202 Gy in coronary angiography, 1.162 Gy in Percutaneus Transluminal Coronary Angioplasty (PTCA) and 0.128 in abdominal angiography. The correlation of DAP and MSD was significant (r = 0.7). The correlation was DAP and fluoroscopy time was also significant (r = 0.8). Conclusion: The highest MSD values were found in PTCA which is a therapeutic procedure. Two PTCAs out of the 57 procedures measured in total had MSD over the threshold of 2 Gy for deterministic effects (MSD 1 = 3.0 Gy and MSD 2 3.3 Gy). The large variations in MSD reveal the need to continuously monitor patient doses in IR and IC procedures with special emphasis in PTCA procedure. (author)

  18. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

    International Nuclear Information System (INIS)

    2015-10-01

    At the beginning of 2014, there were 437 nuclear power reactors in operation and 72 reactors under construction. To date, around 370 500 t (HM) (tonnes of heavy metal) of spent fuel have been discharged from reactors, and approximately 253 700 t (HM) are stored at various storage facilities. Although wet storage at reactor sites still dominates, the amount of spent fuel being transferred to dry storage technologies has increased significantly since 2005. For example, around 28% of the total fuel inventory in the United States of America is now in dry storage. Although the licensing for the construction of geological disposal facilities is under way in Finland, France and Sweden, the first facility is not expected to be available until 2025 and for most States with major nuclear programmes not for several decades afterwards. Spent fuel is currently accumulating at around 7000 t (HM) per year worldwide. The net result is that the duration of spent fuel storage has increased beyond what was originally foreseen. In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. To address this, the IAEA continued the Coordinated Research Project (CRP) on Spent Fuel Performance Assessment and Research (SPAR-III) in 2009 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). This has been achieved through: evaluating surveillance and monitoring programmes of spent fuel and storage facilities; collecting and exchanging relevant experience of spent fuel storage and the impact on associated spent fuel management activities; facilitating the transfer of knowledge by documenting the technical basis for spent fuel storage; creating synergy among research projects of the participating Member States; and developing the capability to assess the impact

  19. Interdisciplinary perspectives on dose limits in radioactive waste management : A research paper developed within the ENTRIA project

    NARCIS (Netherlands)

    Kalmbach, K.; Röhlig, K.-J.

    2016-01-01

    Within the ENTRIA project, an interdisciplinary group of scientists developed a research paper aiming at a synthesis of the technical, sociology of knowledge, legal, societal, and political aspects of dose limits within the field of radioactive waste management. In this paper, the ENTRIA project is

  20. Process, cost modeling and simulations for integrated project development of biomass for fuel and protein

    International Nuclear Information System (INIS)

    Pannir Selvam, P.V.; Wolff, D.M.B.; Souza Melo, H.N.

    1998-01-01

    The construction of the models for biomass project development are described. These models, first constructed using QPRO electronic spread sheet for Windows, are now being developed with the aid of visual and object oriented program as tools using DELPHI V.1 for windows and process simulator SUPERPRO, V.2.7 Intelligent Inc. These models render the process development problems with economic objectives to be solved very rapidly. The preliminary analysis of cost and investments of biomass utilisation projects which are included for this study are: steam, ammonia, carbon dioxide and alkali pretreatment process, methane gas production using anaerobic digestion process, aerobic composting, ethanol fermentation and distillation, effluent treatments using high rate algae production as well as cogeneration of energy for drying. The main project under developments are the biomass valuation projects with the elephant (Napier) grass, sugar cane bagasse and microalgae, using models for mass balance, equipment and production cost. The sensibility analyses are carried out to account for stochastic variation of the process yield, production volume, price variations, using Monte Carlo method. These models allow the identification of economical and scale up problems of the technology. The results obtained with few preliminary project development with few case studies are reported for integrated project development for fuel and protein using process and cost simulation models. (author)

  1. Corrosion database for the nuclear fuel cycle. Sub-project no. 1

    International Nuclear Information System (INIS)

    Schoenfeld, R.; Wegner, K.

    1989-03-01

    The aim of the project was to prepare and process data on corrosion in fuel recycling systems of fast breeder reactors and to store them in a test data base designed as an information system. Based on examinations on the nitric acid corrosion of austenitic steels (typical material/corrosive agent combination used in the reprocessing of burned fuel elements of nuclear power plants) and, in coordination with scientist specialized on materials, the most important characteristics were determined and summarized in a catalogue. This catalogue was realized with the help of a relational data base management system as a scientific data base where the adequate information from the original literature is recorded. (orig./MM) [de

  2. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  3. Projected global health impacts from severe nuclear accidents: Conversion of projected doses to risks on a global scale: Experience from Chernobyl releases

    International Nuclear Information System (INIS)

    Catlin, R.J.; Goldman, M.; Anspaugh, L.R.

    1988-01-01

    Estimates of projected collective dose and average individual dose commitments from Chernobyl releases were made for various regions. Consideration was given to the possible effectiveness of protective actions taken by various countries to reduce projected doses to their populations. Although some preliminary data indicate possible mean reductions of about 25% in total collective doses over the first year, and of about 55% in collective dose to the thyroid, no corrections were made to these dose estimates because of the variable nature of the data. A new combined set of dose-effect models recently published by the United States Nuclear Regulatory Commission was then applied to estimate the ranges of possible future additional health effects due to the Chernobyl accident. In this method possible health effects are estimated on an individual site basis and the results are then summed. Both absolute and relative risk projection models are used. By use of these methods, ''best'' estimates of possible additional health effects were projected for the Northern Hemisphere as follows: 1) over the next 50 years, up to 28 thousand radiation-induced fatal cancers, compared to an expected 600 million cancer deaths from natural or spontaneous causes; 2) over the next year, up to 700 additional cases of severe mental retardation, compared to a normal expectation of 340 thousand cases; and 3) in the first generation, up to 1.9 thousand radiation-induced genetic disorders, compared to 180 million naturally-occurring cases. The possibility of zero health effects at very low doses and dose rates cannot be excluded. Owing to the very large numbers of naturally-occurring health effects, it is unlikely that any additional health effects will be demonstrable except, perhaps, for the more highly exposed population in the immediate vicinity of Chernobyl. 13 refs, 4 figs, 6 tabs

  4. Image quality and radiation dose of low dose coronary CT angiography in obese patients: Sinogram affirmed iterative reconstruction versus filtered back projection

    International Nuclear Information System (INIS)

    Wang, Rui; Schoepf, U. Joseph; Wu, Runze; Reddy, Ryan P.; Zhang, Chuanchen; Yu, Wei; Liu, Yi; Zhang, Zhaoqi

    2012-01-01

    Purpose: To investigate the image quality and radiation dose of low radiation dose CT coronary angiography (CTCA) using sinogram affirmed iterative reconstruction (SAFIRE) compared with standard dose CTCA using filtered back-projection (FBP) in obese patients. Materials and methods: Seventy-eight consecutive obese patients were randomized into two groups and scanned using a prospectively ECG-triggered step-and-shot (SAS) CTCA protocol on a dual-source CT scanner. Thirty-nine patients (protocol A) were examined using a routine radiation dose protocol at 120 kV and images were reconstructed with FBP (protocol A). Thirty-nine patients (protocol B) were examined using a low dose protocol at 100 kV and images were reconstructed with SAFIRE. Two blinded observers independently assessed the image quality of each coronary segment using a 4-point scale (1 = non-diagnostic, 4 = excellent) and measured the objective parameters image noise, signal-to-noise ratio (SNR), and contrast-to-noise ratio (CNR). Radiation dose was calculated. Results: The coronary artery image quality scores, image noise, SNR and CNR were not significantly different between protocols A and B (all p > 0.05), with image quality scores of 3.51 ± 0.70 versus 3.55 ± 0.47, respectively. The effective radiation dose was significantly lower in protocol B (4.41 ± 0.83 mSv) than that in protocol A (8.83 ± 1.74 mSv, p < 0.01). Conclusion: Compared with standard dose CTCA using FBP, low dose CTCA using SAFIRE can maintain diagnostic image quality with 50% reduction of radiation dose.

  5. On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho

    2007-01-01

    In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

  6. Process Intensification in Fuel Cell CHP Systems, the ReforCELL Project

    Directory of Open Access Journals (Sweden)

    José Luis Viviente

    2016-10-01

    Full Text Available This paper reports the findings of a FP7/FCH JU project (ReforCELL that developed materials (catalysts and membranes and an advance autothermal membrane reformer for a micro Combined Heat and Power (CHP system of 5 kWel based on a polymer electrolyte membrane fuel cell (PEMFC. In this project, an active, stable and selective catalyst was developed for the reactions of interest and its production was scaled up to kg scale (TRL5 (TRL: Technology Readiness Level. Simultaneously, new membranes for gas separation were developed. In particular, dense supported thin palladium-based membranes were developed for hydrogen separation from reactive mixtures. These membranes were successfully scaled up to TRL4 and used in lab-scale reactors for fluidized bed steam methane reforming (SMR and autothermal reforming (ATR and in a prototype reactor for ATR. Suitable sealing techniques able to integrate the different membranes in lab-scale and prototype reactors were also developed. The project also addressed the design and optimization of the subcomponents (BoP for the integration of the membrane reformer to the fuel cell system.

  7. The IAEA international project on innovative nuclear reactors and fuel cycles (INPRO): current and future activities

    International Nuclear Information System (INIS)

    Kupitz, J.; Depisch, F.; Kuznetsov, V.

    2004-01-01

    Upon resolutions of the IAEA General Conference in 2000, the IAEA initiated International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). The objective of INPRO, which comprises two phases, is to support sustainable deployment and use of nuclear technology to meet the global energy needs in the next 50 years and beyond. During Phase I, work is subdivided into two sub phases. Phase 1A focused on determining user requirements in the areas of economics, environment, safety, proliferation resistance, and recommendations in the area of so-called crosscutting issues, which are legal, institutional, and infrastructure issues accompanying the deployment of nuclear power, and is targeted at developing a methodology and guidelines for the assessment of various nuclear reactor and fuel cycle concepts and approaches. Phase 1A was finalised in June 2003 with its results now available as IAEA TECDOC-1362. Phase 1B has started in July 2003. During this phase interested Member States are performing case studies to validate the INPRO methodology and, later on, to assess selected innovative nuclear energy systems using the updated INPRO methodology. In accordance with the INPRO Terms of Reference, after successful completion of Phase I, Phase II may be initiated to examine the feasibility of commencing international projects on innovative nuclear energy systems. The paper contains a description of the current and future activities of INPRO and summarizes the outcome of the project.(author)

  8. Fuel Cell Demonstration Project - 200 kW - Phosphoric Acid Fuel Cell Power Plant Located at the National Transportation Research Center: FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Berry, JB

    2005-05-06

    Oak Ridge National Laboratory (ORNL) researches and develops distributed generation technology for the Department of Energy, Energy Efficiency and Renewable Energy Distributed Energy Program. This report describes installation and operation of one such distributed generation system, a United Technology Corporation fuel cell located at the National Transportation Research Center in Knoxville, Tennessee. Data collected from June 2003 to June of 2004, provides valuable insight regarding fuel cell-grid compatibility and the cost-benefit of the fuel cell operation. The NTRC fuel cell included a high-heat recovery option so that use of thermal energy improves project economics and improves system efficiency to 59% year round. During the year the fuel cell supplied a total of 834MWh to the NTRC and provided 300MBtu of hot water. Installation of the NTRC fuel cell was funded by the Distributed Energy Program with partial funding from the Department of Defense's Climate Change Fuel Cell Buy Down Program, administered by the National Energy Technology Laboratory. On-going operational expenses are funded by ORNL's utility budget and are paid from operational cost savings. Technical information and the benefit-cost of the fuel cell are both evaluated in this report and sister reports.

  9. Recalculation with SEACAB of the activation by spent fuel neutrons and residual dose originated in the racks replaced at Cofrentes NPP

    Directory of Open Access Journals (Sweden)

    Ortego Pedro

    2017-01-01

    Full Text Available In order to increase the storage capacity of the East Spent Fuel Pool at the Cofrentes NPP, located in Valencia province, Spain, the existing storage stainless steel racks were replaced by a new design of compact borated stainless steel racks allowing a 65% increase in fuel storing capacity. Calculation of the activation of the used racks was successfully performed with the use of MCNP4B code. Additionally the dose rate at contact with a row of racks in standing position and behind a wall of shielding material has been calculated using MCNP4B code as well. These results allowed a preliminary definition of the burnker required for the storage of racks. Recently the activity in the racks has been recalculated with SEACAB system which combines the mesh tally of MCNP codes with the activation code ACAB, applying the rigorous two-step method (R2S developed at home, benchmarked with FNG irradiation experiments and usually applied in fusion calculations for ITER project.

  10. Recalculation with SEACAB of the activation by spent fuel neutrons and residual dose originated in the racks replaced at Cofrentes NPP

    Science.gov (United States)

    Ortego, Pedro; Rodriguez, Alain; Töre, Candan; Compadre, José Luis de Diego; Quesada, Baltasar Rodriguez; Moreno, Raul Orive

    2017-09-01

    In order to increase the storage capacity of the East Spent Fuel Pool at the Cofrentes NPP, located in Valencia province, Spain, the existing storage stainless steel racks were replaced by a new design of compact borated stainless steel racks allowing a 65% increase in fuel storing capacity. Calculation of the activation of the used racks was successfully performed with the use of MCNP4B code. Additionally the dose rate at contact with a row of racks in standing position and behind a wall of shielding material has been calculated using MCNP4B code as well. These results allowed a preliminary definition of the burnker required for the storage of racks. Recently the activity in the racks has been recalculated with SEACAB system which combines the mesh tally of MCNP codes with the activation code ACAB, applying the rigorous two-step method (R2S) developed at home, benchmarked with FNG irradiation experiments and usually applied in fusion calculations for ITER project.

  11. General-purpose heat source project and space nuclear safety fuels program. Progress report, February 1980

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1980-05-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space Nuclear Safety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work

  12. Basic research in support of innovative fuels design for the Generation IV systems (F-BRIDGE project)

    International Nuclear Information System (INIS)

    Valot, Carole; Bertolus, Marjorie; Konings, Rudy; Somers, Joe; Groot, Sander de

    2010-01-01

    F-BRIDGE (Basic Research in support of Innovative Fuels Design for the GEN IV systems) is a 4-year project which started in 2008. It seeks to bridge the gap between basic research and technological applications for generation IV nuclear reactor systems. One of the challenges for the next generation of reactors is to significantly increase the efficiency in designing innovative fuels. The object of the F-BRIDGE project is to complement the empirical approach by a physically-based description of fuel and cladding materials to enable a rationalization of the design process and a better selection of promising fuel systems. Advanced modelling and separate effects experiments are carried out in order to obtain more exact physical descriptions of ceramic fuels and cladding, at relevant scales from the atomic to the macroscopic scale. Research is also focused on assessing and improving 'sphere-pac' fuel, a composite-ceramics concept which has shown promise. The project activities can be broken down into four main areas: (i) Basic research investigations using a multi-scale approach in both experimentation and modelling to enable the generation of missing basic data, the identification of relevant mechanisms and the development of appropriate models; (ii) Transfer between technological issues and basic research by bringing together within the same project materials scientists, engineers and end-users; (iii) Assessment of the drawbacks and benefits of the sphere-pac fuel application to various Generation IV systems; (iv) Education and training to promote research in the field of fuel materials, to ensure the exchange of results and ideas among the participants and to link the project with other related European or international initiatives. The project relies on the complementary expertise of 19 partners: nuclear and non nuclear research organisations, universities, a nuclear engineering company, as well as technology and project management consultancy small and medium

  13. EcoDoses. Improving radiological assessment of doses to man from terrestrial ecosystems. A status report for the NKS-B project 2004

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Sven P.; Isaksson, M.; Nilsson, Elisabeth (and others)

    2005-07-01

    The NKS B-programme EcoDoses project started in 2003 as a collaboration between all the Nordic countries. The aim of the project is to improve the radiological assessments of doses to man from terrestrial ecosystems. The present report sums up the work performed in the second phase of the project. The main topics in 2004 have been: (i) A continuation of previous work with a better approach for estimating global fallout on a regional or national scale, based on a correlation between precipitation and deposition rates. (ii) Fur-ther extension of the EcoDoses milk database. Estimation of effective ecological half lives of {sup 137}Cs in cows milk focussing on suitable post-Chernobyl time-series. Modelling integrated transfer of {sup 13}7{sup C}s to cow's milk from Nordic countries. (iii) Determination of effective ecological half lives for fresh water fish from Nordic lakes. (iv) Investigate ra-dioecological sensitivity for Nordic populations. (v) Food-chain modelling using the Eco-sys-model, which is the underlying food- and dose-module in several computerised deci-sion-making systems. (au)

  14. EcoDoses. Improving radiological assessment of doses to man from terrestrial ecosystems. A status report for the NKS-B project 2004

    International Nuclear Information System (INIS)

    Nielsen, Sven P.; Isaksson, M.; Nilsson, Elisabeth

    2005-07-01

    The NKS B-programme EcoDoses project started in 2003 as a collaboration between all the Nordic countries. The aim of the project is to improve the radiological assessments of doses to man from terrestrial ecosystems. The present report sums up the work performed in the second phase of the project. The main topics in 2004 have been: (i) A continuation of previous work with a better approach for estimating global fallout on a regional or national scale, based on a correlation between precipitation and deposition rates. (ii) Fur-ther extension of the EcoDoses milk database. Estimation of effective ecological half lives of 137 Cs in cows milk focussing on suitable post-Chernobyl time-series. Modelling integrated transfer of 13 7 C s to cow's milk from Nordic countries. (iii) Determination of effective ecological half lives for fresh water fish from Nordic lakes. (iv) Investigate ra-dioecological sensitivity for Nordic populations. (v) Food-chain modelling using the Eco-sys-model, which is the underlying food- and dose-module in several computerised deci-sion-making systems. (au)

  15. Internal dose assessments: Uncertainty studies and update of ideas guidelines and databases within CONRAD project

    International Nuclear Information System (INIS)

    Marsh, J. W.; Castellani, C. M.; Hurtgen, C.; Lopez, M. A.; Andrasi, A.; Bailey, M. R.; Birchall, A.; Blanchardon, E.; Desai, A. D.; Dorrian, M. D.; Doerfel, H.; Koukouliou, V.; Luciani, A.; Malatova, I.; Molokanov, A.; Puncher, M.; Vrba, T.

    2008-01-01

    The work of Task Group 5.1 (uncertainty studies and revision of IDEAS guidelines) and Task Group 5.5 (update of IDEAS databases) of the CONRAD project is described. Scattering factor (SF) values (i.e. measurement uncertainties) have been calculated for different radionuclides and types of monitoring data using real data contained in the IDEAS Internal Contamination Database. Based upon this work and other published values, default SF values are suggested. Uncertainty studies have been carried out using both a Bayesian approach as well as a frequentist (classical) approach. The IDEAS guidelines have been revised in areas relating to the evaluation of an effective AMAD, guidance is given on evaluating wound cases with the NCRP wound model and suggestions made on the number and type of measurements required for dose assessment. (authors)

  16. Netherlands contribution to the EC project: Benchmark exercise on dose estimation in a regulatory context

    International Nuclear Information System (INIS)

    Stolk, D.J.

    1987-04-01

    On request of the Netherlands government FEL-TNO is developing a decision support system with the acronym RAMBOS for the assessment of the off-site consequences of an accident with hazardous materials. This is a user friendly interactive computer program, which uses very sophisticated graphical means. RAMBOS supports the emergency planning organization in two ways. Firstly, the risk to the residents in the surroundings of the accident is quantified in terms of severity and magnitude (number of casualties, etc.). Secondly, the consequences of countermeasures, such as sheltering and evacuation, are predicted. By evaluating several countermeasures the user can determine an optimum policy to reduce the impact of the accident. Within the framework of the EC project 'Benchmark exercise on dose estimation in a regulatory context' on request of the Ministry of Housing, Physical Planning and Environment calculations were carried out with the RAMBOS system. This report contains the results of these calculations. 3 refs.; 2 figs.; 10 tabs

  17. Dose and risk assessment approach for the Fernald CERCLA D ampersand D Project

    International Nuclear Information System (INIS)

    Throckmorton, J.D.; Clark, T.R.; Waligora, S.J. Jr.; Haaker, R.F.

    1994-01-01

    At the Fernald Environmental Management Project (FEMP) the uranium processing facilities used from the 1952 through 1989 are near or beyond their intended design life. These current conditions present an increasing probability for future releases of hazardous substances to the environment. To support a decision by the U.S. Department of Energy (DOE) and the Environmental Protection Agency (EPA) to remediate the buildings, a dose and risk assessment was performed to determine the extent of exposure that would be associated with the controlled decontamination and dismantlement (D ampersand D) of the Fernald facilities. A conceptual risk assessment model was developed, with exposure mechanisms and associated pathways for each potential receptor. The three receptor groups were defined as: the remediation workers, other on-site workers (those not performing D ampersand D), and off-site residents. For use in the conceptual model, an airborne source term was developed through process knowledge, other historical information and data, and air sample data from within the facilities. Individual and collective doses and risks were developed for each receptor and for each population group. The risk assessment demonstrated that all exposures resulting from the action would be within the acceptable DOE administrative control level of 2.0 rem per year for occupational workers and the acceptable EPA risk range from 10 -6 to 10 -4 for the general public

  18. Cavities at the Si projected range by high dose and energy Si ion implantation in Si

    International Nuclear Information System (INIS)

    Canino, M.; Regula, G.; Lancin, M.; Xu, M.; Pichaud, B.; Ntzoenzok, E.; Barthe, M.F.

    2009-01-01

    Two series of n-type Si samples α and β are implanted with Si ions at high dose (1 x 10 16 ) and high energies, 0.3 and 1.0 MeV, respectively. Both sort of samples are then implanted with 5 x 10 16 He cm -2 (at 10 or 50 keV) and eventually with B atoms. Some of the samples are annealed at temperatures ranging from 800 to 1000 deg. C to allow the thermal growth of He-cavities, located between sample surface and the projected range (R p ) of Si. After the triple ion implantation, which corresponds to defect engineering, samples were characterized by cross-section transmission electron microscopy (XTEM). Voids (or bubbles) are observed not only at the R p (He) on all annealed samples, but also at the R p (Si) on β samples implanted with He at 50 keV. The samples are also studied by positron annihilation spectroscopy (PAS) and the spectra confirm that as-implanted samples contain di-vacancies and that the annealed ones, even at high temperature have bigger open volumes, which are assumed to be the same voids observed by XTEM. It is demonstrated that a sole Si implantation at high energy and dose is efficient to create cavities which are thermally stable up to 1000 deg. C only in the presence of He.

  19. Radioactive Air Emissions Notice of Construction (NOC) for the Solid Waste Treatment Facility (T Plant) Fuel Removal Project

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, R.E.

    2000-11-16

    This NOC describes the activities to remove all spent nuclear fuel (SNF) assemblies from the spent fuel pool in the T Plant Complex 221-T canyon for interim storage in the Canister Storage Building (CSB). The unabated total effective dose equivalent (TEDE) estimated for the public hypothetical maximally exposed individual (MEI) is 5.7 E-6 millirem (mrem) per year for this fuel removal NOC. The abated TEDE conservatively is estimated to account for 2.9 E-9 mrem per year to the MEI.

  20. Optimization in the nuclear fuel cycle I: Temporal variation of dose rate

    International Nuclear Information System (INIS)

    Pereira, W.S.; Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Fernandes, T.S.; Mello, C.R.; Kelecom, A.

    2017-01-01

    Radioprotection aims to protect man and the environment from the harmful effects of radiation. Radioprotection is based on three fundamental principles: justification, dose limitation and optimization. Optimization is a complementary principle to dose limitation and should be applied in all phases of development, and even in unregulated situations. The aim of this work is to use the exposure rate as a tool to optimize radioprotection. The exposure rate at a nuclear facility was monitored at 15 points for one year and statistical tools for data analysis were proposed as auxiliary tools for the process of optimizing the dose rates measured at the facility. A total of 9,125 exposure-rate measures were performed during 2014. The monthly averages were organized by sampling point and by month of the year. No statistical difference was observed in the monthly variation of the dose rate. Therefore, this variable can not be used in the optimization process in this nuclear installation

  1. The ''Ulisse'' project for the treatment of radioactive wastes of ENEA Department of Fuel

    International Nuclear Information System (INIS)

    Broglia, M.

    1988-01-01

    The high-level liquid wastes producedby the Nuclear Fuel Cycle Department of the Italian Committee for R and D of Nuclear and Alternative Energy since 1970 up today, are characterized by a very high aluminum content which would seriously limit the option of their vitrification. A simple chemical process for the separation of the radioactive fraction from the aluminum nitrate solution before vitrification has been proposed, reducing by a factor 10 the finally obtainable glass-waste volume (the ''ULISSE'' Project). This comparative study of the ''ULISSE'' high-level waste and detailed compositions of preferred products from various international Laboratories should supply a good initial condition in a glass matrix design for waste immobilization, and a feed-back control on the Project parameters

  2. Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

    International Nuclear Information System (INIS)

    MITCHELL, R.M.

    1999-01-01

    This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins

  3. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1980-02-01

    Studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of LASL are presented. The three programs involved are: general-purpose heat source development; space nuclear safety; and fuels program. Three impact tests were conducted to evaluate the effects of a high temperature reentry pulse and the use of CBCF on impact performance. Additionally, two 238 PuO 2 pellets were encapsulated in Ir-0.3% W for impact testing. Results of the clad development test and vent testing are noted. Results of the environmental tests are summarized. Progress on the Stirling isotope power systems test and the status of the improved MHW tests are indicated. The examination of the impact failure of the iridium shell of MHFT-65 at a fuel pass-through continued. A test plan was written for vibration testing of the assembled light-weight radioisotopic heater unit. Progress on fuel processing is reported

  4. Externalities of fuel cycles 'ExternE' project. Oil fuel cycle. Estimation of physical impacts and monetary valuation for priority impact pathways

    International Nuclear Information System (INIS)

    Friedrich, R.; Krewitt, W.; Mayerhofer, P.; Trukenmueller, A.; Gressmann, A.; Runte, K.-H.; Kortum, G.; Weltschev, M.

    1994-01-01

    Fuel cycle externalities are the costs imposed on society and the environment that are not accounted for by the producers and consumers of energy. They include damage to health, forests, crops, natural ecosystems and the built environment. Traditional assessment of fuel cycles has ignored these effects and the energy sector is consequently distorted in favor of technologies with significant environmental burdens. Concern over widespread degradation of the environment resulting from fuel cycle emissions has mounted since the late 1960s. In the early 1970s the potential for long range atmospheric transport of certain pollutants was recognized. The effects of acidifying pollutants, ozone precursors and greenhouse gases have caused particular concern. This is reflected in recent trends in economic thought, particularly the emphasis on sustainable development and the use of market mechanisms for environmental regulation. It has thus become increasingly clear that the external impacts of energy use are significant and should be considered by energy planners. Although the theoretical basis for including external costs in decision making processes has been generally agreed, an acceptable methodology for their calculation and integration has not been established. The studies of Hohmeyer (1988), Ottinger et al. (1990) and Friedrich and Voss (1993) provide the background for such work, though they are of a somewhat preliminary nature. We need to improve the methods employed and the quality of models and data used so that planning decisions can be based at least partly on the results. It is particularly important that the site and project specificity of many impacts is recognized. In consequence of this a collaborative project between Directorate General XII (Science, Research and Technology) of the European Commission and the United States Department of Energy has been established to identify the most appropriate methodology for this type of work. The current study has three

  5. Externalities of fuel cycles 'ExternE' project. Lignite fuel cycle. Estimation of physical impacts and monetary valuation for priority impact pathways

    International Nuclear Information System (INIS)

    Friedrich, R.; Krewitt, W.; Mayerhofer, P.; Trukenmueller, A.; Gressmann, A.

    1994-01-01

    Fuel cycle externalities are the costs imposed on society and the environment that are not accounted for by the producers and consumers of energy. They include damage to health, forests, crops, natural ecosystems and the built environment. Traditional assessment of fuel cycles has ignored these effects and the energy sector is consequently distorted in favor of technologies with significant environmental burdens. Concern over widespread degradation of the environment resulting from fuel cycle emissions has mounted since the late 1960s. In the early 1970s the potential for long range atmospheric transport of certain pollutants was recognized. The effects of acidifying pollutants, ozone precursors and greenhouse gases have caused particular concern. This is reflected in recent trends in economic thought, particularly the emphasis on sustainable development and the use of market mechanisms for environmental regulation. It has thus become increasingly clear that the external impacts of energy use are significant and should be considered by energy planners. Although the theoretical basis for including external costs in decision making processes has been generally agreed, an acceptable methodology for their calculation and integration has not been established. The studies of Hohmeyer (1988] and Ottinger et al. [1990] provide the background for such work, though they are of a somewhat preliminary nature [Friedrich, Voss, 1993]. We need to improve the methods employed and the quality of models and data used so that planning decisions can be based at least partly on the results. If is particularly important that the site and project specificity of many impacts is recognized. In consequence of this a collaborative project between Directorate General XII (Science, Research and Technology) of the European Commission and the United States Department of Energy has been established to identify the most appropriate methodology for this type of work. The current study has three

  6. Radiological impact of a spent fuel disposal in a deep geological granite formation - results of the european spa project

    International Nuclear Information System (INIS)

    Baudoin, P.; Gay, D.; Certes, C.; Serres, C.

    2000-01-01

    The SPA project (Spent fuel disposal Performance Assessment) is the latest of four integrated performance assessment exercises on nuclear waste disposal in geological formations, carried out in the framework of the European Community 'Nuclear Fission' Research Programmes. The SPA project, which was undertaken by ENRESA, GRS, IPSN, NRG, SCK.CEN and VTT between May 1996 and April 1999, was devoted to the study of disposal of spent fuel in various host rock formations (clay, crystalline rocks and salt formation). This project is a direct continuation of the efforts made by the European Community since 1982 to build a common understanding of the methods applicable to deep disposal performance assessment. (authors)

  7. ACSEPT a European project for a new step in the future demonstration of advanced fuel processing

    International Nuclear Information System (INIS)

    Bourg, S.; Hill, C.; Caravaca, C.; Espartero, A.; Rhodes, C.; Taylor, R.; Harrison, M.; EKBERG, C.; GEIST, A.; Modolo, G.; Cassayre, L.; Malmbeck, R.; De Angelis, G.; Bouvet, S.; Klaassen, F.

    2010-01-01

    For more than fifteen years, a European scientific community has joined its effort to develop and optimise processes for the partitioning of actinides from fission products. In an international context of 'nuclear renaissance', the upcoming of a new generation of nuclear reactor (Gen IV) will require the development of associated advanced closed fuel cycles which answer the needs of a sustainable nuclear energy: the minimization of the production of long lived radioactive waste but also the optimization of the use of natural resources with an increased resistance to proliferation. Actually, Partitioning and Transmutation (P and T), associated to a multi-recycling of all transuranics (TRUs), should play a key role in the development of this sustainable nuclear energy. By joining together 34 Partners coming from European universities, nuclear research bodies and major industrial players in a multidisciplinary consortium, the FP7 EURATOM-Fission Collaborative Project ACSEPT (Actinide recycling by Separation and Transmutation), started in 2008 for four year duration, provides the sound basis and fundamental improvements for future demonstrations of fuel treatment in strong connection with fuel fabrication techniques. Consistently with potentially viable recycling strategies, ACSEPT therefore provides a structured R and D framework to develop chemical separation processes compatible with fuel fabrication techniques, with a view to their future demonstration at the pilot level. ACSEPT is organized into three technical domains: (i) Considering technically mature aqueous separation processes, ACSEPT works to optimize and select the most promising ones dedicated either to actinide partitioning or to group actinide separation. (ii) Concerning high temperature pyrochemical separation processes, ACSEPT focuses on the enhancement of the two reference cores of process selected within previous projects. R and D efforts are now devoted to key scientific and technical points

  8. Integrated project for increasing the capacity of spent fuel pools at Cofrentes NPP

    International Nuclear Information System (INIS)

    Rebollo Garcia, C.; Arana, S.

    1996-01-01

    The current storage capacity of the Cofrentes NPP will have reached its limit by the end of its 15th cycle, in the year 2005. The works performed by Empresarios Agrupados for IBERDROLA show that it is possible to increase this capacity in successive phases, so as to make the Power Plant self-sufficient for 16 more years (up to 2021) in the case of compact storage, or for 50 more years (2055) in the case of consolidated storage or second level storage. Optimisation of the management of high-activity wastes goes with a series of tasks which come under the group referred to as Integrated Project for Increasing the Capacity of Spent Fuel Pools. The main activities of the project can be summarised under the following three items: increase of storage capacity (feasibility study, specification for the purchase of racks, manufacture, assembly and tests), improvement of the capacity of the pool cooling system and modification of the components and accessories located inside the pools which interfere with the new racks. Another series of activities with less technical and economic impact are: modification of fuel handling machines, management of generated radwaste, licensing and modification of plant documentation (seismic analysis, radiation areas, as-built drawings and verification of the validation of purification and HVAC systems). (Author)

  9. Data for FUMEX: Results from fuel behavior studies at the OECD Halden Reactor Project for model validation and development

    International Nuclear Information System (INIS)

    Wiesenack, W.

    1997-01-01

    Investigations of phenomena associated with extended or high burn-up are an important part of the fuel and materials testing programme carried out at the OECD Halden Reactor Project. The in-core studies comprise long term fuel rod behavior as well as the response to power ramps. Performance is assessed through measurements of fuel centre temperature, rod pressure, elongation of cladding and fuel stack, and cladding diameter changes obtained during full power reactor operation. Data from fuel behavior studies at the OECD Halden Reactor Project, provided for the IAEA co-ordinated research programme FUMEX, are used to elucidate short and long-term developments of fuel behavior. The examples comprise: fuel conductivity degradation manifested as a gradual temperature increase with burn-up; the influence of a combination of small gap/high fission gas release on fuel centre temperature (situation at high burn-up); fission gas release during normal operation and power ramps, and the possibility of a burn-up enhancement; PCMI reflected by cladding elongation, also for the case of a nominally open gap, and the change of interaction onset with burn-up. (author). 10 refs, 9 figs, 1 tab

  10. Data for FUMEX: Results from fuel behavior studies at the OECD Halden Reactor Project for model validation and development

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-08-01

    Investigations of phenomena associated with extended or high burn-up are an important part of the fuel and materials testing programme carried out at the OECD Halden Reactor Project. The in-core studies comprise long term fuel rod behavior as well as the response to power ramps. Performance is assessed through measurements of fuel centre temperature, rod pressure, elongation of cladding and fuel stack, and cladding diameter changes obtained during full power reactor operation. Data from fuel behavior studies at the OECD Halden Reactor Project, provided for the IAEA co-ordinated research programme FUMEX, are used to elucidate short and long-term developments of fuel behavior. The examples comprise: fuel conductivity degradation manifested as a gradual temperature increase with burn-up; the influence of a combination of small gap/high fission gas release on fuel centre temperature (situation at high burn-up); fission gas release during normal operation and power ramps, and the possibility of a burn-up enhancement; PCMI reflected by cladding elongation, also for the case of a nominally open gap, and the change of interaction onset with burn-up. (author). 10 refs, 9 figs, 1 tab.

  11. Site Assessment Report/Corrective Action Plan Fuel Hydrant System (Site SS-41) Project DKFX937081R1, Charleston Air Force Base, South Carolina

    National Research Council Canada - National Science Library

    1996-01-01

    ... #4 and the aircraft apron. Limited investigations were performed on those portions of the fuel hydrant system located on the west side of the drainage ditch due to ongoing Navy fuel hydrant system and aircraft apron renovation projects...

  12. Optimisation of water chemistry to ensure reliable water reactor fuel performance at high burnup and in ageing plant (FUWAC): an International Atomic Energy Agency coordinated research project

    Energy Technology Data Exchange (ETDEWEB)

    Killeen, J.C. [International Atomic Energy Agency, Vienna (Austria); Nordmann, F. [Advanced Nuclear Technology International Europe AB, Beauchamp (France); Schunk, J. [Paks NPP (Hungary); Vonkova, K. [Nuclear Research Inst., Rez (Czech Republic)

    2010-07-01

    The IAEA project 'Optimisation of Water Chemistry to ensure Reliable Water Reactor Fuel Performance at High Burnup and in Aging Plant' (FUWAC) was initiated with the objectives of monitoring, maintaining and optimising water chemistry regimes in primary circuits of water cooled power reactors, taking into account high burnup operation, mixed cores and plant aging, including following issues and remedies. This report provides some highlights of the work undertaken by the project participants. Clad oxidation studies have been undertaken and include operational data from the South Ukraine WWER where no corrosion problems have been seen on either Westinghouse ZIRLO™ or Russian alloy E110 fuel cladding. Work on the Russian alloy E110 showed that potassium in the coolant is preferable to lithium for mitigating fuel cladding oxidation. Studies on crud behaviour in PWR have shown a dependence on crud thickness and pHT. The nature and mechanisms for boron deposition in fuel cladding cruds have been investigated which is the root cause of crud induced power shifts (CIPS). Operational experience at French PWRs shows no difference in the CIPS behaviour between units with Alloy 600 or 690 steam generators, whilst Korean experience provides information on the Ni/Fe ratio on fuel cladding crud and the occurrence of CIPS. Coolant additions have been studied, for example in BWR units using zinc addition, crud is more tenacious. Zinc is also added to PWR units, mainly for dose rate control and in some cases for PWSCC mitigation of Alloy 600. At low levels there has been no clear evidence of any effect of zinc on CIPS, but there is a benefit on fuel oxidation. It is suggested that zinc addition should be considered where there is SG replacement or fuel core management modification. One possibility for the elimination of fuel crud is decontamination. Such an operation is time consuming, expensive, includes several risks of corrosion and induces a large quantity of

  13. Technical Approach and Plan for Transitioning Spent Nuclear Fuel (SNF) Project Facilities to the Environmental Restoration Program

    International Nuclear Information System (INIS)

    SKELLY, W.A.

    1999-01-01

    This document describes the approach and process in which the 100-K Area Facilities are to be deactivated and transitioned over to the Environmental Restoration Program after spent nuclear fuel has been removed from the K Basins. It describes the Transition Project's scope and objectives, work breakdown structure, activity planning, estimated cost, and schedule. This report will be utilized as a planning document for project management and control and to communicate details of project content and integration

  14. Dose analysis on high performance vault storage system of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Kei-ichiro; Maki, Koichi; Shimizu, Masashi; Oda, Masashi; Kumagai, Naoki [Hitachi, Ltd., Power and Industrial Systems R and D Laboratory, Hitachi, Ibaraki (Japan); Hoshikawa, Tadahiro; Oyama, Kenichi; Kanai, Hidetoshi [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    2000-03-01

    The radiation shielding design for the high performance vault storage system is studied in order to satisfy the design targets at the controlled area and the site boundaries. The additional gamma-ray shields in front of the storage tubes and the shielding structures at the entrance of the ducts are installed. The dose rates at the inlet and the outlet of the ducts are estimated by 3D calculation. And the dose rates at the controlled area and the site boundaries are evaluated taking the effect of the direct radiation and the indirect one (skyshine) into consideration. The dose rates at the controlled area and the site boundaries are about 7x10{sup -7} Sv/h and 3x10{sup -10} Sv/h, respectively. Thus, we have the prospect to satisfy the design targets. (author)

  15. The food chain and dose submodel, CALDOS, for the assessment of Canada's nuclear fuel waste management concept

    International Nuclear Information System (INIS)

    Zach, R.; Sheppard, S.C.

    1992-12-01

    The food chain and dose submodel, CALDOS, for assessing Canada's nuclear fuel management (NFWM) concept of disposal in a vault deep in the Canadian Shield is presented. Together with the surface water, soil and atmosphere submodels, CALDOS is integrated into a comprehensive, probabilistic biosphere model for post-closure assessment. This model is representative of the Canadian Shield in Ontario and CALDOS is fully generic. CALDOS calculates radionuclide transfer through the environment to make dose predictions for man. It considers 68 radionuclides explicitly and takes into account another 28 short-lived daughters in the dose calculations. Nine potentially toxic elements are also considered. CALDOS is of the multiplicative chain type for most of the radionuclides, but some, such as 3H , 129I and 222R n, are treated specially. The model accounts for all the major internal exposure pathways, including root uptake, leaf deposition, terrestrial animal's drinking water, terrestrial animal soil ingestion, freshwater fish ingestion, human drinking water, human soil ingestion and human inhalation. External exposure from air immersion, water immersion, ground and building materials are also considered. Dose predictions are based on the recommendations of the International Commission on Radiation Protection (ICRP 26) methodologies, ICRP reference man (ICRP 23) and the critical group concept. CALDOS considers ingrowth of some radioactive daughters, radionuclide availability in soil, recycling and depletion. The model has numerous parameters, some element, radionuclide or food type specific. Sensitivity analysis is used to assess parameter importance in dose prediction. Quality assurance is addressed through general literature, model and parameter evaluations, specifically designed for environmental assessment models. This also involves validation and code comparison studies. (author). 43 refs., 36 tabs., 24 figs

  16. The food chain and dose submodel, CALDOS, for the assessment of Canada`s nuclear fuel waste management concept

    Energy Technology Data Exchange (ETDEWEB)

    Zach, R; Sheppard, S C

    1992-12-01

    The food chain and dose submodel, CALDOS, for assessing Canada`s nuclear fuel management (NFWM) concept of disposal in a vault deep in the Canadian Shield is presented. Together with the surface water, soil and atmosphere submodels, CALDOS is integrated into a comprehensive, probabilistic biosphere model for post-closure assessment. This model is representative of the Canadian Shield in Ontario and CALDOS is fully generic. CALDOS calculates radionuclide transfer through the environment to make dose predictions for man. It considers 68 radionuclides explicitly and takes into account another 28 short-lived daughters in the dose calculations. Nine potentially toxic elements are also considered. CALDOS is of the multiplicative chain type for most of the radionuclides, but some, such as {sup 3H}, {sup 129I} and {sup 222R}n, are treated specially. The model accounts for all the major internal exposure pathways, including root uptake, leaf deposition, terrestrial animal`s drinking water, terrestrial animal soil ingestion, freshwater fish ingestion, human drinking water, human soil ingestion and human inhalation. External exposure from air immersion, water immersion, ground and building materials are also considered. Dose predictions are based on the recommendations of the International Commission on Radiation Protection (ICRP 26) methodologies, ICRP reference man (ICRP 23) and the critical group concept. CALDOS considers ingrowth of some radioactive daughters, radionuclide availability in soil, recycling and depletion. The model has numerous parameters, some element, radionuclide or food type specific. Sensitivity analysis is used to assess parameter importance in dose prediction. Quality assurance is addressed through general literature, model and parameter evaluations, specifically designed for environmental assessment models. This also involves validation and code comparison studies. (author). 43 refs., 36 tabs., 24 figs.

  17. EcoDoses improving radiological assessment of doses to man from terrestrial ecosystems. A status report for the NKS-B project 2003

    Energy Technology Data Exchange (ETDEWEB)

    Bergan, T. [Lavrans Skuterud, Haevard Thoerring (Norway); Liland, A. [Norwegian Radiation Protection Authority (NRPA) (Denmark)] (eds.)

    2004-05-01

    The NKS B-programme EcoDoses project started in 2003 as a collaboration between all the Nordic countries. The aim of the project is to improve the radiological assessments of doses to man from terrestrial ecosystems. The first part, conducted in 2003, has focussed on an extensive collation and review of both published and unpublished data from all the Nordic countries for the nuclear weapons fallout period and the post-Chemobyl period. This included data on radionuclides in air filters, precipitation, soil samples, milk and reindeer. Based on this, an improved model for estimating radioactive fallout based on precipitation data during the nuclear weapons fallout period has been developed. Effective ecological half- lives for 137Cs and 90Sr in milk have been calculated for the nuclear weapons fallout period. For reindeer the ecological half- lives for 137Cs have been calculated for both the nuclear weapons fallout period and the post-Chemobyl period. The data were also used to compare modelling results with observed concentrations. This was done at a workshop where the radioecological food-and-dose module in the ARGOS decision support system was used to predict transfer of deposited radionuclides to foodstuffs and subsequent radiation doses to man. The work conducted the first year is presented in this report and gives interesting, new results relevant for terrestrial radioecology. (au)

  18. Time/motion observations and dose analysis of reactor loading, transportation, and dry unloading of an overweight truck spent fuel shipment

    International Nuclear Information System (INIS)

    Hostick, C.J.; Lavender, J.C.; Wakeman, B.H.

    1992-04-01

    This document presents observed activity durations and radiation dose analyses for an overweight truck shipment of pressurized water reactor (PWR) spent fuel from the Surry Power Station in Virginia to the Idaho National Engineering Laboratory. The shipment consisted of a TN-8L shipping cask carrying three 9-year-old PWR spent fuel assemblies. Handling times and dose analyses for at-reactor activities were completed by Virginia Electric and Power Company (Virginia Power) personnel. Observations of in-transit and unloading activities were made by Pacific Northwest Laboratory (PNL) personnel, who followed the shipment for approximately 2800 miles and observed cask unloading activities. In-transit dose estimates were calculated using dose rate maps provided by Virginia Power for a fully loaded TN-8L shipping cask. The dose analysis for the cask unloading operations is based on the observations of PNL personnel

  19. The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    Juergen Kupitz

    2002-01-01

    This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited 'all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology' (GC(44)/RES/21) and invited Member States to consider to contribute to a task force on innovative nuclear reactors and fuel cycle (GC(44)/RES/22). In response to this invitation, the IAEA initiated an 'International Project on Innovative Nuclear Reactors and Fuel Cycles', INPRO. The Terms of Reference for INPRO were adopted at a preparatory meeting in November 2000, and the project was finally launched by the INPRO Steering Committee in May 2001. At the General Conference in 2001, first progress was reported, and the General Conference adopted a resolution on 'Agency Activities in the Development of Innovative Nuclear Technology' [GC(45)/RES/12, Tab F], giving INPRO a broad basis of support. The resolution recognized the 'unique role that the Agency can play in international collaboration in the nuclear field'. It invited both 'interested Member States to contribute to innovative nuclear technology activities' at the Agency as well as the Agency itself 'to continue it's efforts in these areas'. Additional endorsement came in a UN General Assembly resolution in December 2001 (UN GA 2001, A/RES/56/94), that again emphasized 'the unique role that the Agency can play in developing user requirements and in addressing safeguards, safety and environmental questions for innovative reactors and their fuel cycles' and stressed 'the need for international collaboration in the development of innovative nuclear technology'. As of February 2002, the following countries or entities have become members of INPRO: Argentina

  20. Occupational radiation dose assessment for a non site specific spent fuel storage facility

    International Nuclear Information System (INIS)

    Hadley, J.; Eble, R.G. Jr.

    1997-01-01

    To expedite the licensing process of the non site specific Centralized Interim Storage Facility (CISF) the Department of Energy has completed a phase I CISF Topical Safety Analysis Report (TSAR). The TSAR will be used in licensing the phase I CISF if a site is designated. An occupational radiation does assessment of the facility operations is performed as part of the phase I CISF design. The first phase of the CISF has the capability to receive, transfer, and store SNF in dual-purpose cask/canister systems (DPC's). Currently there are five vendor technologies under consideration. The preliminary dose assessment is based on estimated occupational exposures using traditional power plant ISFSI and transport cask handling processes. The second step in the process is to recommend ALARA techniques to reduce potential exposures. A final dose assessment is completed implementing the ALARA techniques and a review is performed to ensure that the design is in compliance with regulatory criteria. The dose assessment and ALARA evaluation are determined using the following input information: Dose estimates from vendor SAR's; ISFSI experience with similar systems; Traditional methods of operations; Expected CISF cask receipt rates; and feasible ALARA techniques. 5 refs., 1 tab

  1. Evaluation of dose to skin surface contamination in the factory Juzbado of fuel elements

    International Nuclear Information System (INIS)

    Ortiz Trujillo, D.; Agustin Perez Fonseca, A.; Alejandro Fuentes, A.

    2013-01-01

    The aim of this work is previously set a simple calculation methodology applicable to the boundary conditions surrounding the environment where skin contamination may have occurred so that you can evaluate in a simple and fast way the dose that the worker is receiving while enduring such pollution. (Author)

  2. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    2000-01-01

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed

  3. Parameter calculation tool for the application of radiological dose projection codes; Herramienta de calculo de parametros para la aplicacion de codigos de proyeccion de dosis radiologicas

    Energy Technology Data Exchange (ETDEWEB)

    Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

    2016-09-15

    The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)

  4. Summary of the OECD Halden Reactor Project Programme on high burn-up fuel performance relevant for BWRs

    International Nuclear Information System (INIS)

    McGrath, M.A.

    1998-01-01

    The basis for the Halden Reactor Project Programme is presented together with an overview of the content of the programme for the time period 1997-1999. The concept of using both separate effects studies, to determine particular fuel properties, and integral rod behaviour studies of commercial fuel is explained. Each of the items in the programme relevant for BWRs are introduced, with most being discussed in further detail. (author)

  5. Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, Paul J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Slonecker, Bruce D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Steen, Franciska H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-15

    The Office of Material