Expert system for aiding fuel exchange planning

International Nuclear Information System (INIS)

Shiratori, Yoshitake.

1992-01-01

A fuel exchange plan is prepared based on intelligence bases such as a method of zoning a reactor core, a fuel arrangement rule and a procedure for fuel exchange, as well as initial data. Then, the fuel exchange plans are patterned to several groups based on the intelligence bases such as patterning of fuel arrangement. A typical pattern is analyzed by a reactor core analyzing code and allowance or rejection for the analyzed patterns is judged based on intelligence bases regarding life time of the reactor core, thermal limit value of fuels, burning of fuel for irradiation test. The result of the judgement is fed-back as an estimation information and the initial data and the intelligence data are added to prepare a plan again. A similar analysis is also conducted for the plan of the rejected group. The allowance or rejection is judged for all of the plans in the same manner. The intelligence bases are used and the result of the judgement is fed-back as a condition for estimation, thereby enabling to remarkably save labors and time and prepare non-typical exchange plans. (N.H.)

Updated FY12 Ceramic Fuels Irradiation Test Plan

International Nuclear Information System (INIS)

Nelson, Andrew T.

2012-01-01

The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

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

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

Planning fuel-conservative descents in an airline environmental using a small programmable calculator: algorithm development and flight test results

Energy Technology Data Exchange (ETDEWEB)

Knox, C.E.; Vicroy, D.D.; Simmon, D.A.

1985-05-01

A simple, airborne, flight-management descent algorithm was developed and programmed into a small programmable calculator. The algorithm may be operated in either a time mode or speed mode. The time mode was designed to aid the pilot in planning and executing a fuel-conservative descent to arrive at a metering fix at a time designated by the air traffic control system. The speed model was designed for planning fuel-conservative descents when time is not a consideration. The descent path for both modes was calculated for a constant with considerations given for the descent Mach/airspeed schedule, gross weight, wind, wind gradient, and nonstandard temperature effects. Flight tests, using the algorithm on the programmable calculator, showed that the open-loop guidance could be useful to airline flight crews for planning and executing fuel-conservative descents.

Characterization program management plan for Hanford K Basin spent nuclear fuel

International Nuclear Information System (INIS)

Lawrence, L.A.

1998-01-01

The management plan developed to characterize the K Basin Spent Nuclear Fuel was revised to incorporate actions necessary to comply with the Office of Civilian Radioactive Waste Management Quality Assurance Requirements Document 0333P. This plan was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. This revision to the Program Management Plan replaces Westinghouse Hanford Company with Duke Engineering and Services Hanford, Inc., updates the various activities where necessary, and expands the Quality Assurance requirements to meet the applicable requirements document. Characterization will continue to utilize the expertise and capabilities of both organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for Duke Engineering and Services Hanford, Inc. and Pacific Northwest National Laboratory to support the Spent Nuclear Fuels Project at Hanford

Advanced Research Reactor Fuel Development

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

2006-04-15

RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

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

Cost-Effective Fuel Treatment Planning

Science.gov (United States)

Kreitler, J.; Thompson, M.; Vaillant, N.

2014-12-01

The cost of fighting large wildland fires in the western United States has grown dramatically over the past decade. This trend will likely continue with growth of the WUI into fire prone ecosystems, dangerous fuel conditions from decades of fire suppression, and a potentially increasing effect from prolonged drought and climate change. Fuel treatments are often considered the primary pre-fire mechanism to reduce the exposure of values at risk to wildland fire, and a growing suite of fire models and tools are employed to prioritize where treatments could mitigate wildland fire damages. Assessments using the likelihood and consequence of fire are critical because funds are insufficient to reduce risk on all lands needing treatment, therefore prioritization is required to maximize the effectiveness of fuel treatment budgets. Cost-effectiveness, doing the most good per dollar, would seem to be an important fuel treatment metric, yet studies or plans that prioritize fuel treatments using costs or cost-effectiveness measures are absent from the literature. Therefore, to explore the effect of using costs in fuel treatment planning we test four prioritization algorithms designed to reduce risk in a case study examining fuel treatments on the Sisters Ranger District of central Oregon. For benefits we model sediment retention and standing biomass, and measure the effectiveness of each algorithm by comparing the differences among treatment and no treat alternative scenarios. Our objective is to maximize the averted loss of net benefits subject to a representative fuel treatment budget. We model costs across the study landscape using the My Fuel Treatment Planner software, tree list data, local mill prices, and GIS-measured site characteristics. We use fire simulations to generate burn probabilities, and estimate fire intensity as conditional flame length at each pixel. Two prioritization algorithms target treatments based on cost-effectiveness and show improvements over those

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...

The Planning of a Small Pilot Plant for Development Work on Aqueous Reprocessing of Nuclear Fuels

Energy Technology Data Exchange (ETDEWEB)

Sjoeborg, T U; Haeffner, E; Hultgren, Aa

1963-10-15

A shielded volume (42 m{sup 3}) in the hot laboratory at Kjeller, Norway, has been used for the installation of a small pilot plant intended for studies on nuclear fuel reprocessing. During the first period of operation (1963) a plutonium separation method (the Silex process) developed at AB Atomenergi will be studied. This document is a description of the project during the stage of technical planning and chemical process development.

Development and engineering plan for graphite spent fuels conditioning program

International Nuclear Information System (INIS)

Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

1993-09-01

Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

Spent Nuclear Fuel (SNF) Removal Campaign Plan

International Nuclear Information System (INIS)

PAJUNEN, A.L.

2000-01-01

The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage

DOE Hydrogen and Fuel Cells Program Plan (September 2011)

Energy Technology Data Exchange (ETDEWEB)

none,

2011-09-01

The Department of Energy Hydrogen and Fuel Cells Program Plan outlines the strategy, activities, and plans of the DOE Hydrogen and Fuel Cells Program, which includes hydrogen and fuel cell activities within the EERE Fuel Cell Technologies Program and the DOE offices of Nuclear Energy, Fossil Energy, and Science.

Development of nuclear fuel cycle technologies

International Nuclear Information System (INIS)

Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio

1995-01-01

In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)

Plan for Characterization of K Basin Spent Nuclear Fuel (SNF) and Sludge (OCRWM)

International Nuclear Information System (INIS)

TRIMBLE, D.J.

2000-01-01

This is an update of the plan for the characterization of spent nuclear fuel (SNF) and sludge stored in the Hanford K West and K East Basins. The purpose of the characterization program is to provide fuel and sludge data in support of the SNF Project in the effort to remove the fuel from the K Basins and place it into dry storage. Characterization of the K Basin fuel and sludge was initiated in 1994 and has been guided by the characterization plans (Abrefah 1994, Lawrence 1995a, Lawrence 1995b) and the characterization program management plan (PMP) (Lawrence 1995c, Lawrence 1998, Trimble 1999). The fuel characterization was completed in 1999. Summaries of these activities were documented by Lawrence (1999) and Suyama (1999). Lawrence (1999) is a summary report providing a road map to the detailed documentation of the fuel characterization. Suyama (1999) provides a basis for the limited characterization sample size as it relates to supporting design limits and the operational safety envelope for the SNF Project. The continuing sludge characterization is guided by a data quality objective (DQO) (Makenas 2000) and a sampling and analysis plan (SAP) (Baker, Welsh and Makenas 2000) The original intent of the characterization program was ''to provide bounding behavior for the fuel'' (Lawrence 1995a). To accomplish this objective, a fuel characterization program was planned that would provide data to augment data from the literature. The program included in-situ examinations of the stored fuel and laboratory testing of individual elements and small samples of fuel (Lawrence 1995a). Some of the planned tests were scaled down or canceled due to the changing needs of the SNF Project. The fundamental technical basis for the process that will be used to place the K Basin fuel into dry storage was established by several key calculations. These calculations characterized nominal and bounding behavior of fuel in Multi-Canister Overpacks (MCOs) during processing and storage

Fuels planning: science synthesis and integration; social issues fact sheet 01: Developing personal responsibility for fuels reduction: Building a successful program to engage property owners

Science.gov (United States)

Rocky Mountain Research Station USDA Forest Service

2004-01-01

In the course of work as a land manager, you will no doubt be involved in developing programs to achieve various objectives, including the improvement of fuels management on private lands. This fact sheet describes six steps that will help you plan and conduct a successful program.

Emergency planning for fuel cycle facilities

International Nuclear Information System (INIS)

Lacey, L.R.

1991-01-01

In April 1989, NRC published new emergency planning regulations which apply to certain by-product, source, and special nuclear materials licensees including most fuel cycle facilities. In addition to these NRC regulations, other regulatory agencies such as EPA, OSHA, and DOT have regulations concerning emergency planning or notification that may apply to fuel cycle facilities. Emergency planning requirements address such areas as emergency classification, organization, notification and activation, assessment, corrective and protective measures, emergency facilities and equipment, maintaining preparedness, records and reports, and recovery. This article reviews applicable regulatory requirements and guidance, then concentrates on implementation strategies to produce an effective emergency response capability

Proposed plan for critical experiments supporting thorium fuel cycle development

International Nuclear Information System (INIS)

Gore, B.F.

1978-09-01

A preliminary plan is proposed for critical experiments to provide data needed for the recycle of thorium based nuclear fuels. The sequence of experimentation starts with well moderated solutions followed by highly concentrated low moderated solutions. It then progresses through lattices moderated by water, by water plus soluble poisons, and by fissile solutions, to solutions poisoned by raschig rings and soluble poisons. Final experiments would treat lattices moderated by poisoned fissile solution, and arrays of stored fissile units

Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update

International Nuclear Information System (INIS)

1994-09-01

The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF ampersand WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal

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)

CPP-603 Underwater Fuel Storage Facility Site Integrated Stabilization Management Plan (SISMP), Volume I

International Nuclear Information System (INIS)

Denney, R.D.

1995-10-01

The CPP-603 Underwater Fuel Storage Facility (UFSF) Site Integrated Stabilization Management Plan (SISMP) has been constructed to describe the activities required for the relocation of spent nuclear fuel (SNF) from the CPP-603 facility. These activities are the only Idaho National Engineering Laboratory (INEL) actions identified in the Implementation Plan developed to meet the requirements of the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1 to the Secretary of Energy regarding an improved schedule for remediation in the Defense Nuclear Facilities Complex. As described in the DNFSB Recommendation 94-1 Implementation Plan, issued February 28, 1995, an INEL Spent Nuclear Fuel Management Plan is currently under development to direct the placement of SNF currently in existing INEL facilities into interim storage, and to address the coordination of intrasite SNF movements with new receipts and intersite transfers that were identified in the DOE SNF Programmatic and INEL Environmental Restoration and Waste Management Environmental Impact Statement Record, of Decision. This SISMP will be a subset of the INEL Spent Nuclear Fuel Management Plan and the activities described are being coordinated with other INEL SNF management activities. The CPP-603 relocation activities have been assigned a high priority so that established milestones will be meet, but there will be some cases where other activities will take precedence in utilization of available resources. The Draft INEL Site Integrated Stabilization Management Plan (SISMP), INEL-94/0279, Draft Rev. 2, dated March 10, 1995, is being superseded by the INEL Spent Nuclear Fuel Management Plan and this CPP-603 specific SISMP

Update of the Used Fuel Dispositon Campaign Implementation Plan

Energy Technology Data Exchange (ETDEWEB)

McMahon, Kevin A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mackinnon, Robert James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Swift, Peter N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Birkholzer, Jens T. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2014-10-01

This Update to the Used Fuel Disposition Campaign Implementation Plan provides summary level detail describing how the Used Fuel Disposition Campaign (UFDC) supports achievement of the overarching mission and objectives of the Department of Energy Office of Nuclear Energy Fuel Cycle Technologies Program, building on work completed in this area since 2009. This implementation plan begins with the assumption of target dates that are set out in the January 2013 DOE Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (http://energy.gov/downloads/strategy-management-and-disposal-used-nuclearfuel- and-high-level-radioactive-waste). These target dates and goals are summarized in section III. This implementation plan will be maintained as a living document and will be updated as needed in response to available funding and progress in the Used Fuel Disposition Campaign and the Fuel Cycle Technologies Program.

Progress on LEU very high density fuel and target development in Argentina

International Nuclear Information System (INIS)

Balart, S.; Cabot, P.; Calzetta, O.; Duran, A.; Garces, J.; Hermida, J.D.; Manzini, A.; Pasqualini, E.; Taboada, H.

2006-01-01

Since last RRFM meeting, CNEA has continued on new LEU fuel and target development activities. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, to optimize techniques to recover U from silicide scrap samples as cold test for radiowaste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: - Completion of the RA-6 reactor conversion to LEU; - Improvement on fuel development and production facilities to implement new technologies, including NDT techniques to assess bonding quality; - Irradiation of miniplates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; - Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

DOE-owned spent nuclear fuel program plan

International Nuclear Information System (INIS)

1995-11-01

The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines

DOE-owned spent nuclear fuel program plan

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-11-01

The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines.

DOE-owned spent nuclear fuel strategic plan. Revision 1

International Nuclear Information System (INIS)

1996-09-01

The Department of Energy (DOE) is responsible for safely and efficiently managing DOE-owned spent nuclear fuel (SNF) and SNF returned to the US from foreign research reactors (FRR). The fuel will be treated where necessary, packaged suitable for repository disposal where practicable, and placed in interim dry storage. These actions will remove remaining vulnerabilities, make as much spent fuel as possible ready for ultimate disposition, and substantially reduce the cost of continued storage. The goal is to complete these actions in 10 years. This SNF Strategic Plan updates the mission, vision, objectives, and strategies for the management of DOE-owned SNF articulated by the SNF Strategic Plan issued in December 1994. The plan describes the remaining issues facing the EM SNF Program, lays out strategies for addressing these issues, and identifies success criteria by which program progress is measured. The objectives and strategies in this plan are consistent with the following Em principles described by the Assistance Secretary in his June 1996 initiative to establish a 10-year time horizon for achieving most program objectives: eliminate and manage the most serious risks; reduce mortgage and support costs to free up funds for further risk reduction; protect worker health and safety; reduce generation of wastes; create a collaborative relationship between DOE and its regulators and stakeholders; focus technology development on cost and risk reduction; and strengthen management and financial control

Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

International Nuclear Information System (INIS)

Lawrence, L.A.

1995-01-01

A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closed conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford

Critical experiment needs and plans of the consolidated fuel reprocessing program

International Nuclear Information System (INIS)

Primm, R.T.

1984-01-01

An integral part of the United States Department of Energy (DOE) plan for the development of breeder reactors is the development of the capability for fuel reprocessing. The Consolidated Fuel Reprocessing Program (CFRP) was established by the DOE to identify and conduct research and development activities in this area. The DOE is currently proposing that a capability to reprocess fast reactor fuel be established in the Fuels and Materials Examination Facility at the Hanford Engineering Development Laboratory. This capability would include conversion of plutonium nitrate to plutonium oxide. The reprocessing line is designated the Breeder Reprocessing Engineering Test (BRET). Criticality safety remains an important critetion in the design of the BRET. The different steps in the reprocessing are reviewed and areas where additional critical experiments are needed have been indentified as also areas where revision or clarification of existing criticality safety standards are desirable

AECL's progress in DUPIC fuel development

International Nuclear Information System (INIS)

Sullivan, J.D.; Ryz, M.A.; Lee, J.W.

1997-01-01

Previous papers described progress in choosing a fabrication route for the DUPIC (Direct Use of Spent PWR Fuel in CANDU) fuel cycle [1], details of the OREOX (Oxidation Reduction of Oxide fuel) process, and preliminary results of out-cell and small-scale in-cell experiments [2]. AECL's project to develop the DUPIC fuel cycle has now progressed to the stage of fabricating DUPIC fuel elements for irradiation testing in a research reactor. Because of the high radiation fields around the spent PWR fuel, all work is being done in hot cells. The equipment used for fabrication of the DUPIC fuel elements is described in this paper. The commissioning, in-cell installation and current status of the fabrication process are also described and plans for the completion of this phase of the DUPIC project are outlined. The goal of this phase of the project is demonstration of the technical feasibility of the DUPIC fuel cycle. (author)

Development of a strategic plan for an international R and D project on innovative nuclear fuel cycles and power plants

International Nuclear Information System (INIS)

Kendall, J.; Choi, J.S.

2002-01-01

The long-term outlook for nuclear energy should be considered in a broader perspective of future energy needs, operational safety, proliferation and environmental impacts. An Advisory Group Meeting (AGM) on Development of a Strategic Plan for an International R and D Project on Innovative Nuclear Fuel Cycles and Power Plants was convened in Vienna in October 1999 to assess the criteria, the needs for international cooperation, and to formulate a strategic plan for project integration. (author)

Comparison of selected foreign plans and practices for spent fuel and high-level waste management

International Nuclear Information System (INIS)

Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

1990-04-01

This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs

Comparison of selected foreign plans and practices for spent fuel and high-level waste management

Energy Technology Data Exchange (ETDEWEB)

Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

1990-04-01

This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

GRIST-2 preliminary test plan and requirements for fuel fabrication and preirradiation

International Nuclear Information System (INIS)

Tang, I.M.; Harmon, D.P.; Torri, A.

1978-12-01

The preliminary version of the GRIST-2 test plan has been developed for the planned initial 5 years (1984 to 1989) of TREAT-Upgrade in-pile tests. These tests will be employed to study the phenomenology and integral behavior of GCFR core disruptive accidents (CDAs) and to support the Final Safety Analysis Report (FSAR) CDA analyses for the demonstration plant licensing. The preliminary test plan is outlined. Test Phases I and II are for the fresh fuel (preconditioned or not) CDA behavior at the beginning-of-life (BOL) reactor state. Phase III is for the reactor state that contains irradiated fuel with a saturated content of helium and fission gas. Phase IV is for larger bundle tests and scaling effects

Thermal fuel research and development facilities in BNFL

International Nuclear Information System (INIS)

Roberts, V.A.; Vickers, J.

1996-01-01

BNFL is committed to providing high quality, cost effective nuclear fuel cycle services to customers on a National and International level. BNFL's services, products and expertise span the complete fuel cycle; from fuel manufacture through to fuel reprocessing, transport, waste management and decommissioning and the Company maintains its technical and commercial lead by investment in continued research and development (R and D). This paper discusses BNFL's involvement in R and D and gives an account of the current facilities available together with a description of the advanced R and D facilities constructed or planned at Springfields and Sellafield. It outlines the work being carried out to support the company fuel technology business, to (1) develop more cost effective routes to existing fuel products; (2) maximize the use of recycled uranium, plutonium and tails uranium and (3) support a successful MOX business

Background and planning requirements for spent fuel shipments to DOE

Energy Technology Data Exchange (ETDEWEB)

Ravenscroft, Norman [Edlow International Company, 1666 Connecticut Avenue, NW, Suite 201, Washington, DC 20009 (United States)

1996-10-01

Information is provided on the planning required and the factors that must be included in the planning process for spent fuel shipments to DOE. A summary is also provided on the background concerning renewal of the DOE spent fuel acceptance policy in May 1996. (author)

A study on research and development planning of the nuclear energy

International Nuclear Information System (INIS)

Won, Byung Chul; Kim, Y. J.; Hong, J. J.

2002-01-01

In this research planning, planning object is taken as 4 projects in nuclear safety field, 2 projects in reactor and nuclear fuel field, 2 projects in basic and fundamental field among 'Mid and Long-term Nuclear R and D Program'. These projects were all carried out by KAERI and their intermediate R and D phase is closed in 2001. Major planning contents in each project contains 1) R and D's necessity and aim, 2) Technological level and depth, 3) R and D's plan, 4) R and D's results application. The planning results are summarized as follows: - Development of risk management technology - Development of optimal severe accident management strategy and engineering safety features - Development of verification and assessment technology for thermal hydraulic safety - Technology development for enhancing component and structure integrity - Proliferation-resistant fuel technology development - Liquid metal reactor design technology development - Nuclear material technology development: characterization and improvement of nuclear materials - Development of a large proton accelerator for innovative researches

Battery and Fuel Cell Development for NASA's Constellation Missions

Science.gov (United States)

Manzo, Michelle A.

2009-01-01

NASA's return to the moon will require advanced battery, fuel cell and regenerative fuel cell energy storage systems. This paper will provide an overview of the planned energy storage systems for the Orion Spacecraft and the Aries rockets that will be used in the return journey to the Moon. Technology development goals and approaches to provide batteries and fuel cells for the Altair Lunar Lander, the new space suit under development for extravehicular activities (EY A) on the Lunar surface, and the Lunar Surface Systems operations will also be discussed.

Battery and Fuel Cell Development for NASA's Exploration Missions

Science.gov (United States)

Manzo, Michelle A.; Reid, Concha M.

2009-01-01

NASA's return to the moon will require advanced battery, fuel cell and regenerative fuel cell energy storage systems. This paper will provide an overview of the planned energy storage systems for the Orion Spacecraft and the Aries rockets that will be used in the return journey to the Moon. Technology development goals and approaches to provide batteries and fuel cells for the Altair Lunar Lander, the new space suit under development for extravehicular activities (EVA) on the Lunar surface, and the Lunar Surface Systems operations will also be discussed.

1989 basic plan for atomic energy development and utilization

International Nuclear Information System (INIS)

1989-01-01

A Basic Plan for Atomic Energy Development and Utilization has been established each year based on the guidelines set up by the Atomic Energy Commission of Japan, with the aim of promoting the development and utilization of atomic energy schematically and efficiently. The Basic Plan shows specific projects to achieve the objectives specified in the Long-Range Plan for Atomic Energy Development and Utilization. The Basic Plan specifies efforts to be made for overall strengthening of safety measures (safety policies, safety research, disaster prevention, etc.), promotion of nuclear power generation, establishment of the nuclear fuel cycle (securing of uranium, technology for uranium enrichment, reprocessing, etc.), development of new types of power reactors (fast breeder reactor, new types of converter reactors, plutonium fuel processing technology), promotion of leading projects (nuclear fusion, utilization of radiations, atomic powered ships, high-temperature engineering tests), promotion of basic technology development (basic research, training of scientists and engineers), voluntary and active international activities (international cooperation), and acquisition of understanding and cooperation of the general public. (N,K.)

Research and development of Proton-Exchange-Membrane (PEM) fuel cell system for transportation applications. Fuel cell infrastructure and commercialization study

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-11-01

This paper has been prepared in partial fulfillment of a subcontract from the Allison Division of General Motors under the terms of Allison`s contract with the U.S. Department of Energy (DE-AC02-90CH10435). The objective of this task (The Fuel Cell Infrastructure and Commercialization Study) is to describe and prepare preliminary evaluations of the processes which will be required to develop fuel cell engines for commercial and private vehicles. This report summarizes the work undertaken on this study. It addresses the availability of the infrastructure (services, energy supplies) and the benefits of creating public/private alliances to accelerate their commercialization. The Allison prime contract includes other tasks related to the research and development of advanced solid polymer fuel cell engines and preparation of a demonstration automotive vehicle. The commercialization process starts when there is sufficient understanding of a fuel cell engine`s technology and markets to initiate preparation of a business plan. The business plan will identify each major step in the design of fuel cell (or electrochemical) engines, evaluation of the markets, acquisition of manufacturing facilities, and the technical and financial resources which will be required. The process will end when one or more companies have successfully developed and produced fuel cell engines at a profit. This study addressed the status of the information which will be required to prepare business plans, develop the economic and market acceptance data, and to identify the mobility, energy and environment benefits of electrochemical or fuel cell engines. It provides the reader with information on the status of fuel cell or electrochemical engine development and their relative advantages over competitive propulsion systems. Recommendations and descriptions of additional technical and business evaluations that are to be developed in more detail in Phase II, are included.

ArcFuels: an ArcMap toolbar for fuel treatment planning and wildfire risk assessment

Science.gov (United States)

Nicole M. Vaillant; Alan A. Ager

2014-01-01

Fire behavior modeling and geospatial analysis can provide tremendous insight to land managers in defining both the benefits and potential impacts of fuel treatments in the context of land management goals and public expectations. ArcFuels is a streamlined fuel management planning and wildfire risk assessment system that creates a trans-scale (stand to large landscape...

Fuel cell power systems for remote applications. Phase 1 final report and business plan

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

The goal of the Fuel Cell Power Systems for Remote Applications project is to commercialize a 0.1--5 kW integrated fuel cell power system (FCPS). The project targets high value niche markets, including natural gas and oil pipelines, off-grid homes, yachts, telecommunication stations and recreational vehicles. Phase 1 includes the market research, technical and financial analysis of the fuel cell power system, technical and financial requirements to establish manufacturing capability, the business plan, and teaming arrangements. Phase 1 also includes project planning, scope of work, and budgets for Phases 2--4. The project is a cooperative effort of Teledyne Brown Engineering--Energy Systems, Schatz Energy Research Center, Hydrogen Burner Technology, and the City of Palm Desert. Phases 2 through 4 are designed to utilize the results of Phase 1, to further the commercial potential of the fuel cell power system. Phase 2 focuses on research and development of the reformer and fuel cell and is divided into three related, but potentially separate tasks. Budgets and timelines for Phase 2 can be found in section 4 of this report. Phase 2 includes: Task A--Develop a reformate tolerant fuel cell stack and 5 kW reformer; Task B--Assemble and deliver a fuel cell that operates on pure hydrogen to the University of Alaska or another site in Alaska; Task C--Provide support and training to the University of Alaska in the setting up and operating a fuel cell test lab. The Phase 1 research examined the market for power systems for off-grid homes, yachts, telecommunication stations and recreational vehicles. Also included in this report are summaries of the previously conducted market reports that examined power needs for remote locations along natural gas and oil pipelines. A list of highlights from the research can be found in the executive summary of the business plan.

Nuclear Fuel Cycle Technologies: Current Challenges and Future Plans - 12558

Energy Technology Data Exchange (ETDEWEB)

Griffith, Andrew [U.S. Department of Energy, Washington, DC (United States)

2012-07-01

The mission of the Office of Nuclear Energy's Fuel Cycle Technologies office (FCT program) is to provide options for possible future changes in national nuclear energy programs. While the recent draft report of the Blue Ribbon Commission on America's Nuclear Future stressed the need for organization changes, interim waste storage and the establishment of a permanent repository for nuclear waste management, it also recognized the potential value of alternate fuel cycles and recommended continued research and development in that area. With constrained budgets and great expectations, the current challenges are significant. The FCT program now performs R and D covering the entire fuel cycle. This broad R and D scope is a result of the assignment of new research and development (R and D) responsibilities to the Office of Nuclear Energy (NE), as well as reorganization within NE. This scope includes uranium extraction from seawater and uranium enrichment R and D, used nuclear fuel recycling technology, advanced fuel development, and a fresh look at a range of disposal geologies. Additionally, the FCT program performs the necessary systems analysis and screening of fuel cycle alternatives that will identify the most promising approaches and areas of technology gaps. Finally, the FCT program is responsible for a focused effort to consider features of fuel cycle technology in a way that promotes nonproliferation and security, such as Safeguards and Security by Design, and advanced monitoring and predictive modeling capabilities. This paper and presentation will provide an overview of the FCT program R and D scope and discuss plans to analyze fuel cycle options and support identified R and D priorities into the future. The FCT program is making progress in implanting a science based, engineering driven research and development program that is evaluating options for a sustainable fuel cycle in the U.S. Responding to the BRC recommendations, any resulting legislative

Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea

Energy Technology Data Exchange (ETDEWEB)

Yang, Jae Ho; Kim, Hyun Gil; In, Wang Kee; Kim, Weon Ju; Koo, Yang Hyum [KAERI, Daejeon (Korea, Republic of); Lee, Seung Jae [KEPCONF, Daejeon (Korea, Republic of)

2016-05-15

Research on accident tolerant fuels (ATFs) is aimed at developing innovative fuels, which can mitigate or prevent the consequences of accidents. In Korea, innovative concepts are being developed to improve fuel safety and reliability of LWRs during accident events and normal operations. ATF technologies will be developed and commercialized through a sequence of long-lead and extensive activities. The interim milestone for new fuel program is that we would be ready for an irradiation test in commercial reactor by 2021. This presentation deals with the status of ATF development in KOREA and plan to implement new fuel technology successfully in commercial nuclear power plants.

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

The 3rd irradiation test plan of DUPIC fuel

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Song, K. C.; Park, J. H. and others

2001-05-01

The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

Status of the atomized uranium silicide fuel development at KAERI

Energy Technology Data Exchange (ETDEWEB)

Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-08-01

While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

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.

Integrated planning for a fuel industry with emphasis on minimum size to fabricate own fuel

International Nuclear Information System (INIS)

Kondal Rao, N.; Katiyar, H.C.; Rajendran, R.; Sinha, K.K.; Swaminathan, N.; Subramanyam, R.B.; Pande, B.P.; Krishnan, T.S.; Agarwala, G.C.; Chandramouli, V.A.

1977-01-01

The Indian nuclear energy programme is based on the utilization of indigenous resources for the economic generation of power, developing its own know-how. In order to gain time, the first nuclear power station at Tarapur is a turn-key job based on enriched uranium fuel. Taking into consideration the established resources of uranium and thorium in the country, a strategy for nuclear power programme has been drawn up. The first phase is based on natural uranium fuel, the second phase on the recycle of plutonium and conversion of thorium and the third phase is the breeder system based on utilization of U 233 and conversion of thorium. This programme is specially significant for India in view of its vast resources of thorium. After the experience and confidence gained with the manufacture of metallic uranium fuel for the research reactors and about 40 tonnes of fuel for the initial loading of the Rajasthan Reactor, the fuel manufacturing programme within the country has been implemented to meet the entire initial and reload fuel requirements. The plant capacities are small compared to similar activities in developed countries. Further, by planning for an integrated fuel and component manufacturing complex, any draw-back in smaller scale of some of the operations is off-set. At the Nuclear Fuel Complex, set up on the above principles, production plants are in operation for the manufacture of reload fuel for the 400 MW Tarapur station, natural uranium oxide fuel, various zircaloy components such as fuel sheaths, pressure tubes, calandria tubes, channels and various other zircaloy components. Provisions have been made to expand the production facilities as the demand for reload fuel grows. With the facilities provided, the production programme can be diversified to take up the production of fast breeder reactor components of stainless steel and also the blanket thorium elements. The unitary control of all aspects of the manufacture and quality control of different types

Development of CANFLEX fuel fabrication technology

Energy Technology Data Exchange (ETDEWEB)

Kang, M. S.; Choi, C. B.; Park, C. H.; Kwon, W. J.; Kim, C. H.; Kim, B. J.; Koo, C. H.; Cho, D. S.; So, D. Y.; Suh, S. W.; Park, C. J.; Chang, D. H.; Yun, S. H. [KEPCO Nuclear Fuel Company, Taejeon (Korea)

2000-04-01

Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU(CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. One of the improvements in CANDU fuel fabrication technology, and advanced method of Zr-Be brazing was developed. For the formation of Zr-Be alloy, preheating and main heating temperature in the furnace is 700 deg C, 1200 deg C respectively. In order to find an appropriate material for the brazing joints in the CANDU fuel, the composition of Zr based amorphous metals were designed. And, the effect of hydrogen on the mechanical properties of cladding sheath and feasibility of the eddy current test to evaluate quality of end cap weld were also studied for the fundamental research purpose. As a preliminary study to suggest optimal way for the mass production of CANFLEX-NU fuel at KNFC the existing CANDU fuel facilities and fabrication/inspection processes were reviewed. The best way is that the current CANDU facility shall be modified to produce small diametrial CANFLEX elements and a new facility shall be constructed to produce large diametrial CANFLEX fuel elements. 46 refs., 99 figs., 10 tabs. (Author)

Development of a strategy for the management of PBMR spent fuel in South Africa

Energy Technology Data Exchange (ETDEWEB)

Smith, S.W., E-mail: Schalk.Smith@necsa.co.z [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa); Bredell, P.J. [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa)

2010-10-15

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would form an integral part of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

Test plan for K-Basin fuel handling tools

International Nuclear Information System (INIS)

Bridges, A.E.

1995-01-01

The purpose of this document is to provide the test plan and procedures for the acceptance testing of the handling tools enveloped for the removal of an N-Reactor fuel element from its storage canister in the K-Basins storage pool and insertion into the Single fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N-Reactor fuel elements is part of the overall characterization effort. New hand tools were required since previous fuel movement has involved grasping the fuel in a horizontal position. The 305 Building Cold Test Facility will be used to conduct the acceptance testing of the Fuel Handling Tools. Upon completion of this acceptance testing and any subsequent training of operators, the tools will be transferred to the 105 KW Basin for installation and use

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

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

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

Energy Technology Data Exchange (ETDEWEB)

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

1993-03-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

Energy Technology Data Exchange (ETDEWEB)

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

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

Operational report, Formation of the XXVII reactor core, plan of fuel exchange

International Nuclear Information System (INIS)

Martinc, R.

1977-01-01

Plan for fuel exchange for formation of the reactor core No. XXVII is presented. This report includes: the quantity of 80% enriched fuel which is input in the core, description of the fuel 'transfer' through the core within this fuelling scheme. It covers the review of reactor safety operating with the core No. XXVII related to reactivity change, thermal load of the fuel channels and fuel burnup. These data result from the analysis based on the same correlated calculation method which was applied for planning the first regular fuel exchange with 80% enriched fuel (core No. XXVI configuration), which has been approved in february 1977. Based on the enclosed data and the fuel exchange according to the proposed procedure it is expected that the reactor operation with core No. XXVII configuration will be safe [sr

Plan for characterization of K Basin spent nuclear fuel and sludge

International Nuclear Information System (INIS)

Lawrence, L.A.; Marschman, S.C.

1995-06-01

This plan outlines a characterization program that supports the accelerated Path Forward scope and schedules for the Spent Nuclear Fuel stored in the Hanford K Basins. This plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years and is limited to in-situ and laboratory examinations of the spent nuclear fuel and sludge in the K East and K West Basins. The program provides bounding behavior of the fuel, and verification and acceptability for three different sludge disposal pathways. Fuel examinations are based on two shipping campaigns for the K West Basin and one from the K East Basin. Laboratory examinations include physical condition, hydride and oxide content, conditioning testing, and dry storage behavior

Development of nuclear fuel for integrated reactor

Energy Technology Data Exchange (ETDEWEB)

Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

1999-04-01

The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

Development of nuclear fuel for integrated reactor

International Nuclear Information System (INIS)

Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M.

1999-04-01

The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO 2 -based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO 2 -based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

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

1991-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

Energy Technology Data Exchange (ETDEWEB)

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

1991-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

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

1991-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. (author)

Evolutionary developments of advanced PWR nuclear fuels and cladding materials

International Nuclear Information System (INIS)

Kim, Kyu-Tae

2013-01-01

Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

Plan for characterization of K Basin Spent Nuclear Fuel and sludge. Revision 1

International Nuclear Information System (INIS)

Lawrence, L.A.

1995-01-01

This plan outlines a Characterization Program that provides the necessary data to support the Integrated Process Strategy scope and schedules for the Spent Nuclear Fuel (SNF) and sludge stored in the Hanford K Basins. The plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years (i.e., FY 1995 through FY 1998) and is limited to in-situ and laboratory examinations of the SNF and sludge in the K East and K West Basins. In order to assure the scope and schedule of the Characterization Program fully supports the Integrated Process Strategy, key project management has approved the plan. The intent of the program is to provide bounding behavior for the fuel, and acceptability for the transfer of the sludge to the Double Shell Tanks. Fuel examinations are based on two shipping compains from the K West Basin and one from the K East Basin with coincident sludge sampling campaings for the associated canister sludge. Sampling of the basin floor and pit sludge will be conducted independent of the fuel and canister sludge shipping activities. Fuel behavior and properties investigated in the laboratory include physical condition, hydride and oxide content, conditioning testing, oxidation kinetics, and dry storage behavior. These laboratory examinations are expected to provide the necessary data to establish or confirm fuel conditioning process limits and support safety analysis. Sludge laboratory examinations include measurement of quantity and content, measurement of properties for equipment design and recovery process limits and support safety analysis. Sludge laboratory examinations include measurement of quantity and content, measurement of properties for equipment design and recovery precesses, tank farm acceptance, simulant development, measurement of corrosion products, and measurements of drying behavior

Fuel tank integrity research : fuel tank analyses and test plans

Science.gov (United States)

2013-04-15

The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

Energy Technology Data Exchange (ETDEWEB)

Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2015-10-15

The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management.

ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

International Nuclear Information System (INIS)

Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo

2015-01-01

The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management

Recent experience in planning, packaging and preparing non-commercial spent fuel for shipment within the United States

International Nuclear Information System (INIS)

Johnson, P.E.; Shappert, L.B.; Turner, D.W.

1996-01-01

US DOE orders dictate that the aluminium clad fuels now stored at ORNL will be shipped to the Savannah River Site. A number of activities had to be carried out in order to ready the fuel for shipping, including choosing a cask capable of transporting the fuel, repackaging the fuel, developing a transportation plan, identifying the appropriate routes, and carrying out a readiness self assessment. These tasks have been successfully completed and are discussed herein

Metallic fuel development

International Nuclear Information System (INIS)

Walters, L.C.

1987-01-01

Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

Nuclear fuel reprocessing deactivation plan for the Idaho Chemical Processing Plant, Revision 1

International Nuclear Information System (INIS)

Patterson, M.W.

1994-10-01

The decision was announced on April 28, 1992 to cease all United States Department of Energy (DOE) reprocessing of nuclear fuels. This decision leads to the deactivation of all fuels dissolution, solvent extraction, krypton gas recovery operations, and product denitration at the Idaho Chemical Processing Plant (ICPP). The reprocessing facilities will be converted to a safe and stable shutdown condition awaiting future alternate uses or decontamination and decommissioning (D ampersand D). This ICPP Deactivation Plan includes the scope of work, schedule, costs, and associated staffing levels necessary to achieve a safe and orderly deactivation of reprocessing activities and the Waste Calcining Facility (WCF). Deactivation activities primarily involve shutdown of operating systems and buildings, fissile and hazardous material removal, and related activities. A minimum required level of continued surveillance and maintenance is planned for each facility/process system to ensure necessary environmental, health, and safety margins are maintained and to support ongoing operations for ICPP facilities that are not being deactivated. Management of the ICPP was transferred from Westinghouse Idaho Nuclear Company, Inc. (WINCO) to Lockheed Idaho Technologies Company (LITCO) on October 1, 1994 as part of the INEL consolidated contract. This revision of the deactivation plan (formerly the Nuclear Fuel Reprocessing Phaseout Plan for the ICPP) is being published during the consolidation of the INEL site-wide contract and the information presented here is current as of October 31, 1994. LITCO has adopted the existing plans for the deactivation of ICPP reprocessing facilities and the plans developed under WINCO are still being actively pursued, although the change in management may result in changes which have not yet been identified. Accordingly, the contents of this plan are subject to revision

Ontario Hydro's plan for used nuclear fuel

International Nuclear Information System (INIS)

Stevens-Guille, P.D.; Howes, H.A.; Freire-Canosa, J.

1992-01-01

A comprehensive plan for the management of used nuclear fuel has been published by Ontario Hydro. In this paper current practices are discussed and actions leading to disposal in a repository are outlined. Extended storage options are discussed should disposal be delayed

Characterization Program Management Plan for Hanford K Basin Spent Nuclear Fuel (SNF) (OCRWM)

International Nuclear Information System (INIS)

BAKER, R.B.; TRIMBLE, D.J.

2000-01-01

The management plan developed to characterize the K Basin spent nuclear fuel (SNF) and sludge was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the SNF stored at the Hanford K Basins. The plan also addressed necessary characterization for the removal, transport, and storage of the sludge from the Hanford K Basins. This plan was revised in 1999 (i.e., Revision 2) to incorporate actions necessary to respond to the deficiencies revealed as the result of Quality Assurance surveillances and audits in 1999 with respect to the fuel characterization activities. Revision 3 to this Program Management Plan responds to a Worker Assessment resolution determined in Fical Year 2000. This revision includes an update to current organizational structures and other revisions needed to keep this management plan consistent with the current project scope. The plan continues to address both the SNF and the sludge accumulated at K Basins. Most activities for the characterization of the SNF have been completed. Data validation, Office of Civilian Radioactive Waste Management (OCRWM) document reviews, and OCRWM data qualification are the remaining SNF characterization activities. The transport and storage of K Basin sludge are affected by recent path forward revisions. These revisions require additional laboratory analyses of the sludge to complete the acquisition of required supporting engineering data. Hence, this revision of the management plan provides the overall work control for these remaining SNF and sludge characterization activities given the current organizational structure of the SNF Project

Greater utilization of wood residue fuels through improved financial planning

International Nuclear Information System (INIS)

Billings, C.D.; Ziemke, M.C.; Stanford, R.

1991-01-01

Recent events have focused attention on the promotion of greater utilization of biomass fuel. Considerations include the need to reduce increases in global warming and also to improve ground level air quality by limiting the use of fossil fuels. However, despite all these important environmentally related considerations, economics remains the most important factor in the decision process used to determine the feasibility of using available renewable fuels instead of more convenient fossil fuels. In many areas of the Southeast, this decision process involves choosing between wood residue fuels such as bark, sawdust and shavings and presently plentiful natural gas. The primary candidate users of wood residue fuels are industries that use large amounts of heat and electric power and are located near centers of activity in the forest products industry such as sawmills, veneer mills and furniture factories. Given that such facilities both produce wood residues and need large amounts of heat and electricity, it is understandable that these firms are often major users of wood-fired furnaces and boilers. The authors have observed that poor or incomplete financial planning by the subject firms is a major barrier to economic utilization of inexpensive and widely available renewable fuels. In this paper, the authors suggest that wider usage of improved financial planning could double the present modest annual incidence of new commercial wood-fueled installation

High performance fuel technology development

Energy Technology Data Exchange (ETDEWEB)

Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

2012-01-15

{omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

Basic plans of nuclear energy development and utilization for fiscal 1982 (report)

International Nuclear Information System (INIS)

1982-01-01

A report by the Nuclear Safety Commission to the Prime Minister, concerning the basic plans of nuclear energy development and utilization for fiscal 1982, was presented; the NSC has decided on the plans drawn up by the Prime Minister. Nuclear power generation as the nucleus of petroleum substitutes must be developed steadily. For the purpose, nuclear fuel cycle should be established, including the securing of uranium resources, uranium enrichment, fuel reprocessing, and waste management. The contents are as follows: the strengthening of nuclear safety measures, the promotion of nuclear power generation, the establishment of nuclear fuel cycle, the development of advanced types of reactors, the research on nuclear fusion, the research and development of nuclear powered ships, the promotion of radiation utilization, the strengthening of basis for nuclear energy development and utilization, the promotion of international cooperation, the strengthening of safeguard and nuclear material protection measures, fiscal 1982 budgets related to nuclear energy. (Mori, K.)

High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

Energy Technology Data Exchange (ETDEWEB)

None

2014-02-27

EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

Planning the RA reactor fuel exchange - Task 3.08/03; Zadatak 3.08/03 - Planiranje izmene goriva reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Dobrosavljevic, N; Strugar, P [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

From the reactor start-up until January 1963, after reactor operation of 1246 MWd , there was a need for fuel exchange. It was necessary to develop a method for planning the refuelling scheme which would enable economic and most appropriate further operation. The most suitable scheme for refuelling would be continuous exchange, but this cannot be performed since each fuel exchange demand reactor shutdown of approximately 15 days. For that reason a compromise scheme was adopted, meaning the fresh fuel is added after each 60-80 days of operation. This report describes the method for approximate calculation of reactivity increase during exchange of the spent fuel element with the fresh one. Detailed plan for exchange of fuel exchange is given in the conclusions.

Fuels planning: science synthesis and integration; social issues fact sheet 13: Strategies for managing fuels and visual quality

Science.gov (United States)

Christine Esposito

2006-01-01

The public's acceptance of forest management practices, including fuels reduction, is heavily based on how forests look. Fuels managers can improve their chances of success by considering aesthetics when making management decisions. This fact sheet reviews a three-part general strategy for managing fuels and visual quality: planning, implementation, and monitoring...

10 CFR 490.203 - Light Duty Alternative Fueled Vehicle Plan.

Science.gov (United States)

2010-01-01

... Efficiency and Renewable Energy, EE-33, 1000 Independence Ave., SW., Washington, DC 20585, or to such other... 10 Energy 3 2010-01-01 2010-01-01 false Light Duty Alternative Fueled Vehicle Plan. 490.203 Section 490.203 Energy DEPARTMENT OF ENERGY ENERGY CONSERVATION ALTERNATIVE FUEL TRANSPORTATION PROGRAM...

Reference spent fuel and its characteristics for the concept development of a deep geological disposal system

Energy Technology Data Exchange (ETDEWEB)

Kang, C. H.; Choi, J. W.; Ko, W. I.; Lee, Y. M.; Park, J. H.; Hwang, Y. S.; Kim, S. K.

1997-09-01

The total amount of spent fuel arisen from the nuclear power plant to be planned by 2010 at the basis of the long-term power development plan announced by MOTIE (Ministry of Trade, Industry and Energy Resource) in 1995 is estimated to derive the disposal capacity of a deep geological repository is derived. The reference spent fuel whose characteristics could be planned is selected by analysing the characteristic data such as initial enrichment, discharge burnup, geometry, dimension, gross weight, etc. Also isotopic concentration, radioactivity, decay heat, hazard index and radiation intensity of a reference spent fuel are quantitatively identified and summarized in order to apply in the concept developing works of a deep geological disposal system. (author). 12 refs., 24 tabs., 14 figs.

Reference spent fuel and its characteristics for the concept development of a deep geological disposal system

International Nuclear Information System (INIS)

Kang, C. H.; Choi, J. W.; Ko, W. I.; Lee, Y. M.; Park, J. H.; Hwang, Y. S.; Kim, S. K.

1997-09-01

The total amount of spent fuel arisen from the nuclear power plant to be planned by 2010 at the basis of the long-term power development plan announced by MOTIE (Ministry of Trade, Industry and Energy Resource) in 1995 is estimated to derive the disposal capacity of a deep geological repository is derived. The reference spent fuel whose characteristics could be planned is selected by analysing the characteristic data such as initial enrichment, discharge burnup, geometry, dimension, gross weight, etc. Also isotopic concentration, radioactivity, decay heat, hazard index and radiation intensity of a reference spent fuel are quantitatively identified and summarized in order to apply in the concept developing works of a deep geological disposal system. (author). 12 refs., 24 tabs., 14 figs

Development of Dynamic Spent Nuclear Fuel Environmental Effect Analysis Model

International Nuclear Information System (INIS)

Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

2010-07-01

The dynamic environmental effect evaluation model for spent nuclear fuel has been developed and incorporated into the system dynamic DANESS code. First, the spent nuclear fuel isotope decay model was modeled. Then, the environmental effects were modeled through short-term decay heat model, short-term radioactivity model, and long-term heat load model. By using the developed model, the Korean once-through nuclear fuel cycles was analyzed. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. If the disposal starts from 2060, the short-term decay heat of Cs-137 and Sr-90 isotopes are W and 1.8x10 6 W in 2100. Also, the total long-term heat load in 2100 will be 4415 MW-y. From the calculation results, it was found that the developed model is very convenient and simple for evaluation of the environmental effect of the spent nuclear fuel

Development of the Canadian used fuel repository engineered barrier system

Energy Technology Data Exchange (ETDEWEB)

Hatton, C., E-mail: chatton@nwmo.ca [Nuclear Waste Management Organization, Toronto, ON (Canada)

2015-07-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for the safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. In implementing APM, the NWMO is committed to ensure consistency with international best practices in the development of its repository system, including any advances in technology. In 2012, the NWMO undertook an optimization study to look at both the design and manufacture of its engineered barriers. This study looked at current technologies for the design and manufacture of used fuel containers, placement technologies, repository design, and buffer and sealing systems, while taking into consideration the state of the art worldwide in repository design and acceptance. The result of that study is the current Canadian engineered barrier system, consisting of a 2.7 tonne used fuel container with a carbon-steel core, copper-coated surface and welded spherical heads. The used fuel container is encapsulated in a bentonite buffer box at the surface and then transferred underground. Once underground, the used fuel is placed into a repository room which is cut into the rock using traditional drill-and-blast technologies. This paper explains the logic for the selection of the container and sealing system design and the development of innovative technologies for their manufacture including the use of laser welding, cold spray and pulsed-electrodeposition copper coating for the manufacture of the used fuel container, isostatic presses for the production of the one-piece bentonite blocks, and slip-skid technologies for placement into the repository. (author)

Development of the Canadian used fuel repository engineered barrier system

International Nuclear Information System (INIS)

Hatton, C.

2015-01-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for the safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. In implementing APM, the NWMO is committed to ensure consistency with international best practices in the development of its repository system, including any advances in technology. In 2012, the NWMO undertook an optimization study to look at both the design and manufacture of its engineered barriers. This study looked at current technologies for the design and manufacture of used fuel containers, placement technologies, repository design, and buffer and sealing systems, while taking into consideration the state of the art worldwide in repository design and acceptance. The result of that study is the current Canadian engineered barrier system, consisting of a 2.7 tonne used fuel container with a carbon-steel core, copper-coated surface and welded spherical heads. The used fuel container is encapsulated in a bentonite buffer box at the surface and then transferred underground. Once underground, the used fuel is placed into a repository room which is cut into the rock using traditional drill-and-blast technologies. This paper explains the logic for the selection of the container and sealing system design and the development of innovative technologies for their manufacture including the use of laser welding, cold spray and pulsed-electrodeposition copper coating for the manufacture of the used fuel container, isostatic presses for the production of the one-piece bentonite blocks, and slip-skid technologies for placement into the repository. (author)

Acceptance criteria for the evaluation of Category 1 fuel cycle facility physical security plans

Energy Technology Data Exchange (ETDEWEB)

Dwyer, P.A.

1991-10-01

This NUREG document presents criteria developed from US Nuclear Regulatory Commission regulations for the evaluation of physical security plans submitted by Category 1 fuel facility licensees. Category 1 refers to those licensees who use or possess a formula quantity of strategic special nuclear material.

Acceptance criteria for the evaluation of Category 1 fuel cycle facility physical security plans

International Nuclear Information System (INIS)

Dwyer, P.A.

1991-10-01

This NUREG document presents criteria developed from US Nuclear Regulatory Commission regulations for the evaluation of physical security plans submitted by Category 1 fuel facility licensees. Category 1 refers to those licensees who use or possess a formula quantity of strategic special nuclear material

Development of System Engineering Technology for Nuclear Fuel Cycle

International Nuclear Information System (INIS)

Kim, Ho Dong; Kim, Sung Ki; Song, Kee Chan

2010-04-01

This report is aims to establish design requirements for constructing mock-up system of pyroprocess by 2011 to realize long-term goal of nuclear energy promotion comprehensive plan, which is construction of engineering scale pyroprocess integrated process demonstration facility. The development of efficient process for spent fuel and establishment of system engineering technology to demonstrate the process are required to develop nuclear energy continuously. The detailed contents of research for these are as follows; - Design of Mock-up facility for demonstrate pyroprocess, Construction, Approval, Trial run, Performance test - Development of nuclear material accountancy technology for unit processes of pyroprocess and design of safeguards system - Remote operation of demonstrating pyroprocess / Development of maintenance technology and equipment - Establishment of transportation system and evaluation of pre-safety for interim storage system - Deriving and implementation of a method to improve nuclear transparency for commercialization proliferation resistance nuclear fuel cycle Spent fuel which is the most important pending problem of nuclear power development would be reduced and recycled by developing the system engineering technology of pyroprocess facility by 2010. This technology would contribute to obtain JD for the use of spent fuel between the ROK-US and to amend the ROK-US Atomic Energy Agreement scheduled in 2014

Implementing the Bio-fuel Plan while considering water resource protection

International Nuclear Information System (INIS)

2006-01-01

This report recalls the objectives of the 'Bio-fuel Plan', and analyses firstly their implications in terms of cultivated surfaces either by using land fallows, or by substitution to other crops, or by intensification, and secondly, the consequences of these different options for water resources. The authors finally discusses the agronomic issue related to the protection of colza, the content of the environmental charter for the cultivation of winter colza, and some financial and practical conditions for the development of energetic crops

NASA's Planned Fuel Cell Development Activities for 2009 and Beyond in Support of the Exploration Vision

Science.gov (United States)

Hoberecht, Mark A.

2010-01-01

NASA s Energy Storage Project is one of many technology development efforts being implemented as part of the Exploration Technology Development Program (ETDP), under the auspices of the Exploration Systems Mission Directorate (ESMD). The Energy Storage Project is a focused technology development effort to advance lithium-ion battery and proton-exchange-membrane fuel cell (PEMFC) technologies to meet the specific power and energy storage needs of NASA Exploration missions. The fuel cell portion of the project has as its focus the development of both primary fuel cell power systems and regenerative fuel cell (RFC) energy storage systems, and is led by the NASA Glenn Research Center (GRC) in partnership with the Johnson Space Center (JSC), the Jet Propulsion Laboratory (JPL), the Kennedy Space Center (KSC), academia, and industrial partners. The development goals are to improve stack electrical performance, reduce system mass and parasitic power requirements, and increase system life and reliability.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

International Nuclear Information System (INIS)

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-01-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

Integrated fuel processor development

International Nuclear Information System (INIS)

Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

2001-01-01

The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

Preliminary plan for decommissioning - repository for spent nuclear fuel

International Nuclear Information System (INIS)

Hallberg, Bengt; Tiberg, Liselotte

2010-06-01

The final disposal facility for spent nuclear fuel is part of the KBS-3 system, which also consists of a central facility for interim storage and encapsulation of the spent nuclear fuel and a transport system. The nuclear fuel repository will be a nuclear facility. Regulation SSMFS 2008:1 (Swedish Radiation Safety Authority's regulations on safety of nuclear facilities) requires that the licensee must have a current decommissioning plan throughout the facility lifecycle. Before the facility is constructed, a preliminary decommissioning plan should be reported to the Swedish Radiation Safety Authority. This document is a preliminary decommissioning plan, and submitted as an attachment to SKB's application for a license under the Nuclear Activities Act to construct, own and operate the facility. The final disposal facility for spent nuclear fuel consists of an above ground part and a below ground part and will be built near Forsmark and the final repository for radioactive operational waste, SFR. The parts above and below ground are connected by a ramp and several shafts, e.g. for ventilation. The below ground part consists of a central area, and several landfill sites. The latter form the repository area. The sealed below ground part constitutes the final repository. The decommissioning is taking place after the main operation has ended, that is, when all spent nuclear fuel has been deposited and the deposition tunnels have been backfilled and plugged. The decommissioning involves sealing of the remaining parts of the below ground part and demolition of above ground part. When decommissioning begins, there will be no contamination in the facility. The demolition is therefore performed as for a conventional plant. Demolition waste is sorted and recycled whenever possible or placed in landfill. Hazardous waste is managed in accordance with current regulations. A ground investigation is performed and is the basis for after-treatment of the site. The timetable for the

Development of a strategy for the management of PBMR spent fuel in South Africa - HTR2008-58047

International Nuclear Information System (INIS)

Smith, S. W.; Bredell, P. J.; Meyer, W. C. M. H.

2008-01-01

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development. (authors)

A Specific Long-Term Plan for Management of U.S. Nuclear Spent Fuel

International Nuclear Information System (INIS)

Levy, Salomon

2006-01-01

A specific plan consisting of six different steps is proposed to accelerate and improve the long-term management of U.S. Light Water Reactor (LWR) spent nuclear fuel. The first step is to construct additional, centralized, engineered (dry cask) spent fuel facilities to have a backup solution to Yucca Mountain (YM) delays or lack of capacity. The second step is to restart the development of the Integral Fast Reactor (IFR), in a burner mode, because of its inherent safety characteristics and its extensive past development in contrast to Acceleration Driven Systems (ADS). The IFR and an improved non-proliferation version of its pyro-processing technology can burn the plutonium (Pu) and minor actinides (MA) obtained by reprocessing LWR spent fuel. The remaining IFR and LWR fission products will be treated for storage at YM. The radiotoxicity of that high level waste (HLW) will fall below that of natural uranium in less than one thousand years. Due to anticipated increased capital, maintenance, and research costs for IFR, the third step is to reduce the required number of IFRs and their potential delays by implementing multiple recycles of Pu and Neptunium (Np) MA in LWR. That strategy is to use an advanced separation process, UREX+, and the MIX Pu option where the role and degradation of Pu is limited by uranium enrichment. UREX+ will decrease proliferation risks by avoiding Pu separation while the MIX fuel will lead to an equilibrium fuel recycle mode in LWR which will reduce U. S. Pu inventory and deliver much smaller volumes of less radioactive HLW to YM. In both steps two and three, Research and Development (R and D) is to emphasize the demonstration of multiple fuel reprocessing and fabrication, while improving HLW treatment, increasing proliferation resistance, and reducing losses of fissile material. The fourth step is to license and construct YM because it is needed for the disposal of defense wastes and the HLW to be generated under the proposed plan. The

As-Built Verification Plan Spent Nuclear Fuel Canister Storage Building MCO Handling Machine

International Nuclear Information System (INIS)

SWENSON, C.E.

2000-01-01

This as-built verification plan outlines the methodology and responsibilities that will be implemented during the as-built field verification activity for the Canister Storage Building (CSB) MCO HANDLING MACHINE (MHM). This as-built verification plan covers THE ELECTRICAL PORTION of the CONSTRUCTION PERFORMED BY POWER CITY UNDER CONTRACT TO MOWAT. The as-built verifications will be performed in accordance Administrative Procedure AP 6-012-00, Spent Nuclear Fuel Project As-Built Verification Plan Development Process, revision I. The results of the verification walkdown will be documented in a verification walkdown completion package, approved by the Design Authority (DA), and maintained in the CSB project files

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

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

A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

International Nuclear Information System (INIS)

Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

New JMTR irradiation test plan on fuels and materials

International Nuclear Information System (INIS)

Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi

2009-01-01

In order to maintain and enhance safety of light water reactors (LWRs) in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency (JAEA) have decided to refurbish the Japan Materials Testing Reactor (JMTR) and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of newdesign fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the Nuclear Safety Research Reactor (NSRR) and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses. (author)

Development of a regional capacity expansion plan in the Russian Federation. Application of the WASP Model

International Nuclear Information System (INIS)

Chernilin, Yu.; Kononov, S.; Zakharova, E.; Kagramanyan, V.; Malenkov, A.

1997-01-01

The Wien Automatic System Planning Package (WASP) is used for the development of optimal capacity expansion plans in Russia. The object of the WASP study is the Central power pool, which is the largest power pool in Russia and has an essential share of nuclear power in electricity generation. The objective of the study is to assess the long-term competitiveness of nuclear power in the region. The major features of the power system analyzed with WASP are the following: 1) four types of electricity generators are considered: condensity fossil fuel plants, cogeneration fossil fuel plants, nuclear power plants and hydraulic plants; 2) nine fuel categories are considered: gas/fuel oil fuel, several types of coal and several nuclear fuels; 3) escalation of capital, operation and maintenance, and fuel costs as a result of economic transition is explicitly modeled. Under these assumptions, a regional optimal capacity expansion plan is developed that showed the following: (a) Until 2004 there is no need for new electricity generation capacities due to the drop in demand in the 90s, certain lifetime margin of existing capacities, committed additions of co-generators and planned refurbishment/repowering measures; (b) The structure of the optimal capacity mix confirms that nuclear power can retain its role as one of the major electricity generation sources in the region. The most important factor with a positive of effect upon the competitiveness of nuclear power plants is the projected escalation of the prices of fossil fuels; (c) The application of WASP has proved that the model can serve as a valuable planning tool at the power pool level in Russia. (author). 14 refs, 8 figs, 10 tabs

Development of a regional capacity expansion plan in the Russian Federation. Application of the WASP Model

Energy Technology Data Exchange (ETDEWEB)

Chernilin, Yu; Kononov, S; Zakharova, E [Russian Research Inst. ` ` Kurchatov Inst.` ` , Moscow (Russian Federation); Kagramanyan, V; Malenkov, A [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1997-09-01

The Wien Automatic System Planning Package (WASP) is used for the development of optimal capacity expansion plans in Russia. The object of the WASP study is the Central power pool, which is the largest power pool in Russia and has an essential share of nuclear power in electricity generation. The objective of the study is to assess the long-term competitiveness of nuclear power in the region. The major features of the power system analyzed with WASP are the following: 1) four types of electricity generators are considered: condensity fossil fuel plants, cogeneration fossil fuel plants, nuclear power plants and hydraulic plants; 2) nine fuel categories are considered: gas/fuel oil fuel, several types of coal and several nuclear fuels; 3) escalation of capital, operation and maintenance, and fuel costs as a result of economic transition is explicitly modeled. Under these assumptions, a regional optimal capacity expansion plan is developed that showed the following: (a) Until 2004 there is no need for new electricity generation capacities due to the drop in demand in the 90s, certain lifetime margin of existing capacities, committed additions of co-generators and planned refurbishment/repowering measures; (b) The structure of the optimal capacity mix confirms that nuclear power can retain its role as one of the major electricity generation sources in the region. The most important factor with a positive of effect upon the competitiveness of nuclear power plants is the projected escalation of the prices of fossil fuels; (c) The application of WASP has proved that the model can serve as a valuable planning tool at the power pool level in Russia. (author). 14 refs, 8 figs, 10 tabs.

Integrated National Energy Planning (INEP) in developing countries

International Nuclear Information System (INIS)

Munasinghe, M.

1989-01-01

Issues of coordinated energy planning are emphasized, with particular reference to interrelationships among the policies adopted in various energy sub-sectors such as electric power (including hydro, nuclear, geothermal, oil and coal sources), petroleum, natural gas, coal, non-conventional (solar, bio-gas, mini-hydro) and traditional fuels (woodfuel, bagasse or vegetable residue). The scope and objectives of integrated national energy planning, the policy tools available, and constraints particular to the developing countries are discussed next. Section 3.0 outlines how energy planning is carried out, while the problems of implementing the resulting policy conclusions are examined in section 4.0. 5 refs, 4 figs

Test plan for spent fuel cladding containment credit tests

International Nuclear Information System (INIS)

Wilson, C.N.

1983-11-01

Lawrence Livermore National Laboratory has chosen Westinghouse Hanford Company as a subcontractor to assist them in determining the requirements for successful disposal of spent fuel rods in the proposed Nevada Test Site repository. An initial scoping test, with the objective of determining whether or not the cladding of a breached fuel rod can be given any credit as an effective barrier to radionuclide release, is described in this test plan. 8 references, 2 figures, 4 tables

Savannah River Site FY 1998 Spent Nuclear Fuel Interim Management Plan

International Nuclear Information System (INIS)

Dupont, M.E.

1998-01-01

This document has been prepared to present in one place the near and long-term plans for safe management of Savannah River Site (SRS) spent nuclear fuel inventories until final disposition has been identified and implemented. The activities described are consistent with FY 1998 Annual Operational Plan guidance and with the December 1997 SRS Accelerated Cleanup Plan update. Summarized are highlights, key decision dates, and baseline assumptions of this plan

Fuels and fire in land-management planning. Part 1. Forest-fuel classification.

Science.gov (United States)

Wayne G. Maxwell; Franklin R. Ward

1981-01-01

This report describes a way to collect and classify the total fuel complex within a planning area. The information can be used as input for appraising and rating probable fire behavior and calculating expected costs and losses from various land uses and management alternatives, reported separately as Part 2 and Part 3 of this series. This total package can be used...

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rader, Jordan D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examination facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

International Nuclear Information System (INIS)

Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-01-01

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

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

Environmental development plan. LWR commercial waste management

International Nuclear Information System (INIS)

1980-08-01

This Environmental Development Plan (EDP) identifies the planning and managerial requirements and schedules needed to evaluate and assess the environmental, health and safety (EH and S) aspects of the Commercial Waste Management Program (CWM). Environment is defined in its broadest sense to include environmental, health (occupational and public), safety, socioeconomic, legal and institutional aspects. This plan addresses certain present and potential Federal responsibilities for the storage, treatment, transfer and disposal of radioactive waste materials produced by the nuclear power industry. The handling and disposal of LWR spent fuel and processed high-level waste (in the event reprocessing occurs) are included in this plan. Defense waste management activities, which are addressed in detail in a separate EDP, are considered only to the extent that such activities are common to the commercial waste management program. This EDP addresses three principal elements associated with the disposal of radioactive waste materials from the commercial nuclear power industry, namely Terminal Isolation Research and Development, Spent Fuel Storage and Waste Treatment Technology. The major specific concerns and requirements addressed are assurance that (1) radioactivity will be contained during waste transport, interim storage or while the waste is considered as retrievable from a repository facility, (2) the interim storage facilities will adequately isolate the radioactive material from the biosphere, (3) the terminal isolation facility will isolate the wastes from the biosphere over a time period allowing the radioactivity to decay to innocuous levels, (4) the terminal isolation mode for the waste will abbreviate the need for surveillance and institutional control by future generations, and (5) the public will accept the basic waste management strategy and geographical sites when needed

First fuel re-load of Angra-1 reactor - Inspection and hearing plan

International Nuclear Information System (INIS)

Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

1985-01-01

The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

Energy Technology Data Exchange (ETDEWEB)

Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

Basic planning of a newly built exclusive ship for spent fuel transport

International Nuclear Information System (INIS)

Obara, I.; Sasao, T.; Akiyama, H.; Kybota, T.

1998-01-01

A commercial reprocessing plant is under construction at the Fuel Cycle Facilities in Rokkasho-mura, Aomori Prefecture. To prepare for the transport of spent nuclear fuels (SF) from all Japanese nuclear power stations to this reprocessing plant, the need for an exclusive transport ship was recognized. Nuclear Fuel Transport Co. Ltd. (NFT), in cooperation with electric power utilities planned the construction of such a ship over a period of several years. During this period NFT developed new types of cask to transport high burn-up spent fuels to the reprocessing plant. Six kinds of casks were developed and 40 units are now under fabrication. The ship was designed to carry a maximum of 20 units. Based on the Irradiated Nuclear Fuel (INF) Code adopted by the International Maritime Organization (IMO), the Japanese Ministry of Transport (MoT) issued new domestic regulations in September, 1995 which covered design criteria for ships carrying Irradiated Nuclear Fuels. The new SF transport ship is the first one to which this new regulation was applied. Although the ship will only ply the coastal routes of Japan, she has been designed to conform with all the international requirements for the Class-3 of the INF Code. In May 1995, Nuclear Fuel Shipping Co. Ltd (NFS), a wholly-owned subsidiary of NFT, concluded a contract with Mitsui Engineering and Shipbuilding Co., Ltd. for the construction of the exclusive transport ship. The keel was laid in November 1995. The ship was launched in april 1996 and named 'Rokuei-Maru'. At the end of September, she was completed and delivered to the ship owner, NFS. (authors)

Fuels planning: science synthesis and integration; economic uses fact sheet 02: log hauling cost

Science.gov (United States)

Rocky Mountain Research Station USDA Forest Service

2004-01-01

Knowing the cost of fuel reduction treatments and associated activities, such as hauling cut trees, is essential for fire and fuels planning. This fact sheet explores the main factors that determine the cost of hauling cut trees and points the user to an interactive tool that can help plan for those and other expenses.

Plan for gas overcrowding; Development in last decade

International Nuclear Information System (INIS)

Unidad de Planeacion Minero Energetica, UPME

2000-01-01

The plan for gas overcrowding, is one of the most important determinations of energy politics in last decade. It is an integral plan of consumption of combustible gas; that is to say, cover the natural gas and the liquefied gas of petroleum. In this article, the UPME wants to show the results of the implementation of this measure. The analysis shows, in the first place the impact achieved in the diversification of the energetic basket, with the penetration of the combustible gas. Facts and indicators of development of the plan are given that include an evaluation of the direct financial benefit that receive the homes, when using a more economic fuel. On the other hand, it puts on of present the government effort carried out through ECOPETROL and that of some private agents to start the plan. Finally, the investments made until today and the problems of low use of some nets of transport, related with the development of the market

Feasibility study on the development of advanced LWR fuel technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs

Feasibility study on the development of advanced LWR fuel technology

Energy Technology Data Exchange (ETDEWEB)

Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

1997-07-01

Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

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

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.

Alcohol fuels for developing countries

International Nuclear Information System (INIS)

Bhattacharya, Partha

1993-01-01

The importance of alcohol as an alternative fuel has been slowly established. In countries such as Brazil, they are already used in transport and other sectors of economy. Other developing countries are also trying out experiments with alcohol fuels. Chances of improving the economy of many developing nations depends to a large extent on the application of this fuel. The potential for alcohol fuels in developing countries should be considered as part of a general biomass-use strategy. The final strategies for the development of alcohol fuel will necessarily reflect the needs, values, and conditions of the individual nations, regions, and societies that develop them. (author). 5 refs

The development of fuel elements for boiling water reactors

International Nuclear Information System (INIS)

Holzer, R.; Kilian, P.

1984-01-01

The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

Fuel Cell Manufacturing Research and Development | Hydrogen and Fuel Cells

Science.gov (United States)

| NREL Fuel Cell Manufacturing Research and Development Fuel Cell Manufacturing Research and Development NREL's fuel cell manufacturing R&D focuses on improving quality-inspection practices for high costs. A researcher monitoring web-line equipment in the Manufacturing Laboratory Many fuel cell

The French UMo group contribution to new LEU fuel development

International Nuclear Information System (INIS)

Hamy, J.M.; Lemoine, P.; Huet, F.; Jarousse, C.; Emin, J.L.

2005-01-01

The French UMo Group was based on a close collaboration between CEA and AREVA's companies strongly involved in the MTR field. The aim of this program was to deliver industrially a high performance LEU UMo fuel able to be reprocessed, and suitable for a wide range of Research Reactor, covering the expected needs for MTR next generation. Since 1999, the program has been focused on industrial aspects with the intention to deal with the whole fuel cycle: manufacturing, irradiation behaviour, fuel characterisation, code development and reprocessing validation. It has been based on the fabrication of full-sized U-7%Mo fuel plates with a density up to 8 gU/cm 3 . The dedicated and advanced R and D means provided by the CEA have been used intensively with the contribution of HFR and BR2 facilities in Europe. This paper presents a synthesis of the program and the corresponding significant results obtained. These results have played a major role as regards the UMo dispersion fuel qualification route by issuing, for the first time, evidence of severe performance limitations. Consequently, the global international effort to develop and qualify a high density LEU UMo fuel has been definitively re-routed and forced to overcome these discrepancies by exploring new technical solutions. A French extended program sustained by a CEA and CERCA collaboration has been launched in 2004 in order to develop a suitable UMo fuel solution. UMo dispersion and monolithic fuel are both investigated through three new full-sized plate irradiations planned in OSIRIS. (author)

Nuclear fuel cycle facilities in the world (excluding the centrally planned economies)

International Nuclear Information System (INIS)

1979-01-01

Information on the existing, under construction and planned fuel cycle facilities in the various countries is presented. Some thirty countries have activities related to different nuclear fuel cycle steps and the information covers the capacity, status, location, and the names of owners of the facilities

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

Development of innovative inspection tools for higher reliability of PHWR fuel

International Nuclear Information System (INIS)

Kamalesh Kumar, B.; Viswanathan, B.; Laxminarayana, B.; Ganguly, C.

2003-01-01

'Full text:' Advent of Computer aided manufacturing systems has led to very high rate of production with greater reliability. The conventional inspection tools and systems, which are often manual based do not complement with output of highly automated production line. In order to overcome the deficiency, a strategic plan was developed for having automated inspection facility for PHWR fuel assembly line. Laser based systems with their inherently high accuracy and quick response times are a favorite for metrology purpose. Non-contact nature of laser-based measurement ensures minimal contamination, low wear and tear and good repeatability. So far two laser-based systems viz. Pellet density measurement systems and triangulation sensors have been developed. Laser based fuel pellet inspection system and PHWR fuel bundle metric station are under development. Machine vision-based systems have been developed to overcome certain limitations when inspection has to be carried out on such a large scale manually. These deficiencies arise from limitations of resolution, accessibility, fatigue and absence of quantification ability. These problems get further compounded in inspection of fuel components because of their relatively small sizes, close tolerances required and the reflective surfaces. PC based vision system has been developed for inspecting components and fuel assemblies. The paper would touch upon the details of the various laser systems and vision systems that have been indigenously developed for PHWR Fuel Metrology and their impact on the assembly production line. (author)

Facility effluent monitoring plan for the 300 Area Fuels Fabrication Facility

International Nuclear Information System (INIS)

Nickels, J.M.; Brendel, D.F.

1991-11-01

A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring system by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The Fuel Fabrication Facility in the Hanford 300 Area supported the production reactors from the 1940's until they were shut down in 1987. Prior to 1987 the Fuel Fabrication Facility released both airborne and liquid radioactive effluents. In January 1987 the emission of airborne radioactive effluents ceased with the shutdown of the fuels facility. The release of liquid radioactive effluents have continued although decreasing significantly from 1987 to 1990

Logistic Fuel Processor Development

National Research Council Canada - National Science Library

Salavani, Reza

2004-01-01

The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

Development of Chemical Technology in Nuclear Fuel Cycle

International Nuclear Information System (INIS)

Jee, Kwang Yong; Kim, W. H.; Kim, J. S.

2007-06-01

This project mainly concentrates on the development of technologies related to elemental analysis for the mass balance of pyro-chemical process, on the development of in-line measurement system for high temperature molten salt, and on the development of radiation shielded LA-ICP-MS and micro-XRD system to evaluate the integrity of nuclear fuel. Chemical analysis methods for the quantitative determination of fissile elements, minor actinide elements, fission products, chemical additive and corrosion products in Uranium Metal Ingots are established. It will be applied to the evaluation of mass balance in electrolytic reduction process for the optimization of the process. Optical fiber based UV-VIS spectrophotometer combined with reaction cell was developed for the measurement of reactions in high temperature molten salt. This system is applicable to in-line monitoring of electro-refining process and contribute to clarify the chemical reactions. Radiation shielded LA-ICP-MS and micro-XRD systems are planned to be used for the analysis of isotopic distribution and structural changes from core to rim of spent nuclear fuel pellet, respectively. The developed techniques can contribute to produce database needed for authorization and practical use of ultra high burn-up fuel. In addition, it can be applicable to the other industries such as microelectronics, nano material science and semiconductor to analyze micro region

Strategic alliances for the development of fuel cell vehicles

Energy Technology Data Exchange (ETDEWEB)

Maruo, Kanehira [Goeteborg Univ. (Sweden). Section of Science and Technology Studies

1998-12-01

The aim of this paper is to explore and describe the current stage of fuel cell vehicle development in the world. One can write three possible future scenarios - an optimistic, a realistic, and a pessimistic scenario: - The optimistic scenario -- The Daimler/Ballard/Ford alliance continues to develop fuel cell stacks and fuel cell vehicle systems as eagerly as they have been doing in recent years. Daimler(/Chrysler)-Benz continues to present its Necar 4, Necar 5, and so on, as planned, and thus keeps Toyota and Honda under severe pressure. Toyota`s and Honda`s real motivation seems to be not to allow Daimler-Benz to be the first to market. Their investment in fuel cell technology will be very large. At the same time, governments and other stake-holders will quickly and in a timely fashion build up infrastructures. We will then see many fuel cell vehicles by 2004. A paradigm shift in automotive technology will have taken place. - The realistic scenario -- Fuel cell vehicles will reach the same level of development by 2004/2005 as pure electric vehicles were at in 1997/1998. This means that fuel cell vehicles will be produced at the rate of several hundred vehicles per year per manufacturer and cost about $40,000 or more, which is still considerably more expensive than ordinary gasoline cars. These fuel cell vehicles will have a performance similar to today`s advanced electric vehicles, e.g., Toyota`s RAV4/EV and Honda`s EV Plus. To go further from this stage to the mass-production stage strong government incentives will be needed. - The pessimistic scenario -- It turns out that fuel cells are not as pure or efficient as in theory and in laboratory experiments. Prices of gasoline and diesel gas continue to be very low. The Californian 10% ZEV Requirement that has been meant to be valid at least ten years from 2003 through 2012 will be suspended or greatly modified. Daimler-Benz, Toyota, and Honda slow down their fuel cell vehicle development activities. No one is

Advanced safeguards research and development plan with an emphasis on its impact on nuclear power-plant design

International Nuclear Information System (INIS)

Tobin, S.J.; Demuth, S.F.; Miller, M.C.; Swinhoe, M.T.; Thomas, K.E.

2007-01-01

One tool for reducing the concern of nuclear proliferation is enhanced safeguards. Present safeguards have evolved over the past 40 years, and future safeguards will grow from this strong base to implement new technologies for improving our ability to quantify nuclear material. This paper will give an overview of the advanced technology research and development plan for safeguarding. One of the research facilities planned by the Department of Energy is the Advanced Fuel Cycle Facility (AFCF), to develop a novel nuclear fuel recycling program. Since the Advanced Fuel Cycle Facility will receive and reprocess spent fuel and will fabricate fast-reactor fuel, a wide breadth of safeguards technologies is involved. A fundamental concept in safeguards is material control and accounting (MCA). 4 topics concerning MCA and requiring further research have been identified: 1) measuring spent fuel, 2) measuring the plutonium content in the electro-refiner with pyro-processing, 3) measuring plutonium in the presence of other actinides, and 4) measuring neptunium and americium in the presence of other actinides. As for the long-term research and development plan for the AFCF, it will include improving MCA techniques as well as introducing new techniques that are not related to MCA, for example, enhanced containment and surveillance, or enhanced process monitoring. The top priority will stay quantifying the plutonium as accurately as possible and to reach this purpose 4 relevant technologies have been identified: 1) the microcalorimeter, 2) the passive neutron-albedo reactivity, 3) list-mode data acquisition, and 4) a liquid-scintillator multiplicity counter. Incorporating safeguards into the initial design of AFCF (safeguards by design) is a central concept. As the technology research and development plan for the Advanced Fuel Cycle Facility is examined, particular attention will be given to safeguards technologies that may affect the physical design of nuclear power plants

Modeling fuel succession

Science.gov (United States)

Brett Davis; Jan van Wagtendonk; Jen Beck; Kent van Wagtendonk

2009-01-01

Surface fuels data are of critical importance for supporting fire incident management, risk assessment, and fuel management planning, but the development of surface fuels data can be expensive and time consuming. The data development process is extensive, generally beginning with acquisition of remotely sensed spatial data such as aerial photography or satellite...

Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

Energy Technology Data Exchange (ETDEWEB)

Benoit, Timothy [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2017-07-05

This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.

Development of the nuclear ship MUTSU spent fuel shipping cask

International Nuclear Information System (INIS)

Ishizuka, M.; Umeda, M.; Nawata, Y.; Sato, H.; Honami, M.; Nomura, T.; Ohashi, M.; Higashino, A.

1989-01-01

After the planned trial voyage (4700 MWD/MTU) of the nuclear ship MUTSU in 1990, her spent fuel assemblies, initially made of two types of enriched UO 2 (3.2wt% and 4.4wt%), will be transferred to the reprocessing plant soon after cooling down in the ship reactor for more than one year. For transportation, the MUTSU spent fuel shipping casks will be used. Prior to transportation to the reprocessing plant, the cooled spent fuel assemblies will be removed from the reactor to the shipping casks and housed at the spent fuel storage facility on site. In designing the MUTSU spent fuel shipping cask, considerations were given to make the leak-tightness and integrity of the cask confirmable during storage. The development of the cask and the storage function demonstration test were performed by Japan Atomic Energy Research Institute (JAERI) and Mitsubishi Heavy Industries, Ltd. (MHI). One prototype cask for the storage demonstration test and licensed thirty-five casks were manufactured between 1987 and 1988

LEU fuel development at CERCA

International Nuclear Information System (INIS)

Durand, Jean Pierre; Ottone, J.C.; Mahe, M.; Ferraz, G.

1998-01-01

The aim of this paper is to detail the recent progress on both U 3 Si 2 high loaded fuels and new γ phase fuels. Concerning high density density silicide plates up to 6 g Ut/cm 3 , the CEA irradiation programme is completed. Data are still under analysis but one can state that the behaviour was globally similar to conventional fuels known in SILOE and OSIRIS reactors. From the new γ fuel point of view, after demonstration feasibility in 1997 of U Mo thermally stable plates loaded up to 8.3 g Ut/cm3, CERCA has analysed the technical ability of quality inspection means assuming that is of an utmost interest for the insurance of a proper use of high performances fuel in reactors. There are mainly two differences between U Mo fuels (and more generally γ fuels) and conventional ones. Firstly, X-ray diffraction analysis on the fuel powder are needed because the chemical analysis is not sufficient to characterise the γ structure requested. Secondly, the physical limits of the Ultrasonic inspection have been reached due to transitory effect between the meat and the edges. Therefore this technic can not applied in the transitory areas. From that knowledge, the manufacture specifications for a plate dedicated to an irradiation plan can be discussed with a clearer view of the main differences with the U 3 Si 2 fuel reference. (author)

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

Remote maintenance system technology development for nuclear fuel cycle plants

International Nuclear Information System (INIS)

Kashihara, Hidechiyo

1984-01-01

The necessity of establishing the technology of remote maintenance, the kinds of maintenance techniques and the change, the image of a facility adopting remote maintenance canyon process, and the outline of the R and D plan to put remote maintenance canyon process in practical use are described. As the objects of development, there are twin arm type servo manipulator system, rack system, remote tube connectors, solution sampling system, inspection system for in-cell equipment, and large plugs for wall penetration. The outline of those are also reported. The development of new remote maintenance technology has been forwarded in the Tokai Works aiming at the application to a glass solidification pilot plant and a FBR fuel recycling test facility. The lowering of the rate of utilization of cells due to poor accessibility and the increase of radiation exposure of workers must be overcome to realize nuclear fuel cycle technology. The maintenance technology is classified into crane canyon method, direct maintenance cell method, remote maintenance cell method and remote maintenance canyon method, and those are described briefly. The development plan of remote maintenance technology is outlined. (Kako, I.)

Automotive Stirling engine development program. [fuel economy assessment

Science.gov (United States)

Kitzner, E. W.

1978-01-01

The Ford/DOE automotive Stirling engine development program is directed towards establishing the technological and developmental base that would enable a decision on whether an engineering program should be directed at Stirling engine production. The fuel economy assessment aims to achieve, with a high degree of confidence, the ERDA proposal estimate of 20.6 MPG (gasoline) for a 4500 lb 1WC Stirling engine passenger car. The current M-H fuel economy projection for the 170 HP Stirling engine is 15.7 MPG. The confidence level for this projection is 32%. A confidence level of 29% is projected for a 22.1 MPG estimate. If all of the planned analyses and test work is accomplished at the end of the one year effort, and the projected improvements are substantiated, the confidence levels would rise to 59% for the 20.6 MPG projection and 54% for the 22.1 MPG projection. Progress achieved thus far during the fuel economy assessment is discussed.

Nuclear fuels technologies: Thermally induced gallium removal system (TIGRS), fiscal year 1998 research and development test plan

International Nuclear Information System (INIS)

Buksa, J.J.; Butt, D.P.; Chidester, K.; DeMuth, S.F.; Havrilla, G.J.; James, C.A.; Kolman, D.G.

1997-01-01

This document details the research and development (R and D) activities that will be conducted in Fiscal Year 1998 (FY98) by the Thermally Induced Gallium Removal System (TIGRS) team for the Department of Energy Office of Fissile Materials Disposition. This work is a continuation and extension of experimental activities that have been conducted in support of using weapons-derived plutonium in the fabrication of mixed-oxide (MOX) nuclear fuel for reactor-based plutonium disposition. The ultimate purpose of this work is to demonstrate adequate Thermally Induced Gallium Removal with a prototypic system. This Test Plan presents more than the FY98 R and D efforts in order to frame the Task in its entirety. To achieve the TIGRS Program objectives, R and D activities during the next two years will be focused on (1) process development leading to a prototypic TIGRS design, and (2) prototypic TIGRS design and testing leading to and including a prototypic demonstration of TIGRS operation. Both the process development and system testing efforts will consist of a series of surrogate-based cold tests and plutonium-based hot tests. Some of this testing has already occurred and will continue into FY99

National Fire Plan Research and Development 2004-2005 accomplishment report

Science.gov (United States)

Michael W. Hilbruner; Paul Keller

2007-01-01

This report highlights accomplishments achieved by USDA Forest Service National Fire Plan Research and Development projects from 2004 through 2005 in four key areas: firefighting, rehabilitation and restoration, hazardous fuels reduction, and community assistance. These highlights illustrate the broad range of knowledge and tools introduced and generated by the...

Strategies for fuel cell product development. Developing fuel cell products in the technology supply chain

International Nuclear Information System (INIS)

Hellman, H.L.

2004-01-01

Due to the high cost of research and development and the broad spectrum of knowledge and competences required to develop fuel cell products, many product-developing firms outsource fuel cell technology, either partly or completely. This article addresses the inter-firm process of fuel cell product development from an Industrial Design Engineering perspective. The fuel cell product development can currently be characterised by a high degree of economic and technical uncertainty. Regarding the technology uncertainty: product-developing firms are more often then not unfamiliar with fuel cell technology technology. Yet there is a high interface complexity between the technology supplied and the product in which it is to be incorporated. In this paper the information exchange in three current fuel cell product development projects is analysed to determine the information required by a product designer to develop a fuel cell product. Technology transfer literature suggests that transfer effectiveness is greatest when the type of technology (technology uncertainty) and the type of relationship between the technology supplier and the recipient are carefully matched. In this line of thinking this paper proposes that the information required by a designer, determined by the design strategy and product/system volume, should be met by an appropriate level of communication interactivity with a technology specialist. (author)

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...

Development of PEM fuel cell technology at international fuel cells

Energy Technology Data Exchange (ETDEWEB)

Wheeler, D.J.

1996-04-01

The PEM technology has not developed to the level of phosphoric acid fuel cells. Several factors have held the technology development back such as high membrane cost, sensitivity of PEM fuel cells to low level of carbon monoxide impurities, the requirement to maintain full humidification of the cell, and the need to pressurize the fuel cell in order to achieve the performance targets. International Fuel Cells has identified a hydrogen fueled PEM fuel cell concept that leverages recent research advances to overcome major economic and technical obstacles.

Planning of fuel coal imports using a mixed integer programming method

Energy Technology Data Exchange (ETDEWEB)

Shih, L.H. [National Cheng Kung University, Tainan (Taiwan). Dept. of Mineral and Petroleum Engineering

1997-12-31

In the public utility and commercial fuel industries, commodities from multiple supply sources are sometimes blended before use to reduce costs and assure quality. A typical example of these commodities is the fuel coal used in coal fired power plants. The diversity of the supply sources for these plants makes the planning and scheduling of fuel coal logistics difficult, especially for a power company that has more than one power plant. This study proposes a mixed integer programming model that provides planning and scheduling of coal imports from multiple suppliers for the Taiwan Power Company. The objective is to minimize total inventory cost by minimizing procurement cost, transportation cost and holding cost. Constraints on the system include company procurement policy, power plant demand, harbor unloading capacity, inventory balance equations, blending requirements, and safety stock. An example problem is presented using the central coal logistics system of the Taiwan Power Company to demonstrate the validity of the proposed model.

Planning of fuel coal imports using a mixed integer programming method

International Nuclear Information System (INIS)

Shih, L.H.

1997-01-01

In the public utility and commercial fuel industries, commodities from multiple supply sources are sometimes blended before use to reduce costs and assure quality. A typical example of these commodities is the fuel coal used in coal fired power plants. The diversity of the supply sources for these plants makes the planning and scheduling of fuel coal logistics difficult, especially for a power company that has more than one power plant. This study proposes a mixed integer programming model that provides planning and scheduling of coal imports from multiple suppliers for the Taiwan Power Company. The objective is to minimize total inventory cost by minimizing procurement cost, transportation cost and holding cost. Constraints on the system include company procurement policy, power plant demand, harbor unloading capacity, inventory balance equations, blending requirements, and safety stock. An example problem is presented using the central coal logistics system of the Taiwan Power Company to demonstrate the validity of the proposed model

Conceptual development of a test facility for spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

Conceptual development of a test facility for spent fuel management

International Nuclear Information System (INIS)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

1997-01-01

Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

Fuels planning: science synthesis and integration; forest structure and fire hazard fact sheet 03: visualizing forest structure and fuels

Science.gov (United States)

Rocky Mountain Research Station USDA Forest Service

2004-01-01

The software described in this fact sheet provides managers with tools for visualizing forest and fuels information. Computer-based landscape simulations can help visualize stand and landscape conditions and the effects of different management treatments and fuel changes over time. These visualizations can assist forest planning by considering a range of management...

SP-100 coated-particle fuel development. Phase I. Final report

International Nuclear Information System (INIS)

1983-03-01

This document is the final report of Phase I of the SP-100 Coated-Particle Fuel Development Program conducted by GA Technologies Inc. for the US Department of Energy under contract DE-AT03-82SF11690. The general objective of the study conducted between September and December 1982 was to evaluate coated-particle type fuel as an alternate or backup fuel to the UO 2 tile-and-fin arrangement currently incorporated into the reference design of the SP-100 reactor core. This report presents and discusses the following topics in the order listed: the need for an alternative fuel for the SP-100 nuclear reactor; an abbreviated description of the reference and coated-particle fuel module concepts; the bases and results of the study and analysis leading to the preliminary design of a coated particle suitable for the SP-100 space power reactor; incorporation of the fuel particles into compacts and heat-pipe-cooled modules; initial efforts and plans to fabricate coated-particle fuel and fuel compacts; the design and performance of the proposed alternative core relative that of the reference fuel; and a summary of critical issues and conclusions consistent with the level of effort and duration of the study

Development of nuclear fuel. Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

1991-07-01

In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

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

Development of Integrity Evaluation Technology for the Long-term Spent Fuel Dry Storage System (1st year Report)

International Nuclear Information System (INIS)

Choi, Jong Won; Kook, Dong Hak; Kim, Jun Sub

2010-05-01

Korea has operated 16 Pressurized Water Reactors(PWR) and has a plan to construct additional nuclear power reactors as only PWR. This causes a big issue of PWR spent fuel accumulation problem now and in the future. KRMC(Korea Radioactive waste Management Coorporation) which was established in 2009 is charged with managing all kinds of radioactive waste that is produced in Korea. KRMC is considering spent fuel dry storage as an option to solve this spent fuel problem and developing the related engineering techniques. KAERI(Korea Atomic Energy Research Institute) also participated in this development and focused on evaluating the spent fuel dry storage system integrity for a long term operation. This report is the first year research product. The aims of the first year work scope are surveying and analyzing models which could anticipate degradation phenomena of the all dry storage components(spent fuel, structure materials, and equipment materials) and selecting items of the tests which are planned to perform in the next project stage. The major work areas consist of 'spent fuel degradation evaluation model development', 'test senario development', 'long-term evaluation of structural material characteristics', and 'dry storage system structure degradation model development'. These works were successfully achieved. This report is expected to contribute for the second year work which includes degradation model development and test senario development, and next project stage

Preliminary plan for decommissioning - repository for spent nuclear fuel; Preliminaer plan foer avveckling - slutfoervar foer anvaent kaernbraensle

Energy Technology Data Exchange (ETDEWEB)

Hallberg, Bengt; Tiberg, Liselotte (Studsvik Nuclear AB, Nykoeping (Sweden))

2010-06-15

The final disposal facility for spent nuclear fuel is part of the KBS-3 system, which also consists of a central facility for interim storage and encapsulation of the spent nuclear fuel and a transport system. The nuclear fuel repository will be a nuclear facility. Regulation SSMFS 2008:1 (Swedish Radiation Safety Authority's regulations on safety of nuclear facilities) requires that the licensee must have a current decommissioning plan throughout the facility lifecycle. Before the facility is constructed, a preliminary decommissioning plan should be reported to the Swedish Radiation Safety Authority. This document is a preliminary decommissioning plan, and submitted as an attachment to SKB's application for a license under the Nuclear Activities Act to construct, own and operate the facility. The final disposal facility for spent nuclear fuel consists of an above ground part and a below ground part and will be built near Forsmark and the final repository for radioactive operational waste, SFR. The parts above and below ground are connected by a ramp and several shafts, e.g. for ventilation. The below ground part consists of a central area, and several landfill sites. The latter form the repository area. The sealed below ground part constitutes the final repository. The decommissioning is taking place after the main operation has ended, that is, when all spent nuclear fuel has been deposited and the deposition tunnels have been backfilled and plugged. The decommissioning involves sealing of the remaining parts of the below ground part and demolition of above ground part. When decommissioning begins, there will be no contamination in the facility. The demolition is therefore performed as for a conventional plant. Demolition waste is sorted and recycled whenever possible or placed in landfill. Hazardous waste is managed in accordance with current regulations. A ground investigation is performed and is the basis for after-treatment of the site. The timetable

Development of a computerized nuclear materials control and accounting system for a fuel reprocessing plant

International Nuclear Information System (INIS)

Crawford, J.M.; Ehinger, M.H.; Joseph, C.; Madeen, M.L.

1979-07-01

A computerized nuclear materials control and accounting system (CNMCAS) for a fuel reprocessing plant is being developed by Allied-General Nuclear Services at the Barnwell Nuclear Fuel Plant. Development work includes on-line demonstration of near real-time measurement, measurement control, accounting, and processing monitoring/process surveillance activities during test process runs using natural uranium. A technique for estimating in-process inventory is also being developed. This paper describes development work performed and planned, plus significant design features required to integrate CNMCAS into an advanced safeguards system

Environmental Development Plan (EDP): space applications

International Nuclear Information System (INIS)

1978-04-01

This Environmental Development Plan (EDP) identifies and examines the environmental, safety, health, and socioeconomic (ES and H) issues associated with the ongoing DOE Space Applications Program and the associated research, development, and demonstration programs. The EDP presents an ES and H research and development (R/D) program and strategy for timely resolutions of the issues and satisfaction of the associated requirements necessary for precluding impediments to the space nuclear systems technology. The EDP has been limited to the consideration of: (1) space nuclear power system nuclear fuel fabrication; (2) space nuclear power system heat source fabrication; (3) testing of subsystems and assembled systems; (4) R and D in support of space nuclear system development; (5) nuclear system responses to launch and reentry accidents; and (6) nuclear system environmental behavior and recovery

Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

Energy Technology Data Exchange (ETDEWEB)

NONE

1985-07-01

Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

International Nuclear Information System (INIS)

1985-01-01

Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

Characterization plan for Hanford spent nuclear fuel

International Nuclear Information System (INIS)

Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

1994-12-01

Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF

FY2015 ceramic fuels development annual highlights

Energy Technology Data Exchange (ETDEWEB)

Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

2015-09-22

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

FY2016 Ceramic Fuels Development Annual Highlights

Energy Technology Data Exchange (ETDEWEB)

Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-01-24

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

Automotive Fuel Processor Development and Demonstration with Fuel Cell Systems

Energy Technology Data Exchange (ETDEWEB)

Nuvera Fuel Cells

2005-04-15

The potential for fuel cell systems to improve energy efficiency and reduce emissions over conventional power systems has generated significant interest in fuel cell technologies. While fuel cells are being investigated for use in many applications such as stationary power generation and small portable devices, transportation applications present some unique challenges for fuel cell technology. Due to their lower operating temperature and non-brittle materials, most transportation work is focusing on fuel cells using proton exchange membrane (PEM) technology. Since PEM fuel cells are fueled by hydrogen, major obstacles to their widespread use are the lack of an available hydrogen fueling infrastructure and hydrogen's relatively low energy storage density, which leads to a much lower driving range than conventional vehicles. One potential solution to the hydrogen infrastructure and storage density issues is to convert a conventional fuel such as gasoline into hydrogen onboard the vehicle using a fuel processor. Figure 2 shows that gasoline stores roughly 7 times more energy per volume than pressurized hydrogen gas at 700 bar and 4 times more than liquid hydrogen. If integrated properly, the fuel processor/fuel cell system would also be more efficient than traditional engines and would give a fuel economy benefit while hydrogen storage and distribution issues are being investigated. Widespread implementation of fuel processor/fuel cell systems requires improvements in several aspects of the technology, including size, startup time, transient response time, and cost. In addition, the ability to operate on a number of hydrocarbon fuels that are available through the existing infrastructure is a key enabler for commercializing these systems. In this program, Nuvera Fuel Cells collaborated with the Department of Energy (DOE) to develop efficient, low-emission, multi-fuel processors for transportation applications. Nuvera's focus was on (1) developing fuel

International Source Book: Nuclear Fuel Cycle Research and Development Vol 1 Volume 1

Energy Technology Data Exchange (ETDEWEB)

Harmon, K. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lakey, L. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

1983-07-01

This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This first volume includes the overview and the program summaries of those countries listed alphabetically from Argentina to Italy.

Canadian fuel development program

International Nuclear Information System (INIS)

Gacesa, M.; Young, E.G.

1992-11-01

CANDU power reactor fuel has demonstrated an enviable operational record. More than 99.9% of the bundles irradiated have provided defect-free service. Defect excursions are responsible for the majority of reported defects. In some cases research and development effort is necessary to resolve these problems. In addition, development initiatives are also directed at improvements of the current design or reduction of fueling cost. The majority of the funding for this effort has been provided by COG (CANDU Owners' Group) over the past 10 to 15 years. This paper contains an overview of some key fuel technology programs within COG. The CANDU reactor is unique among the world's power reactors in its flexibility and its ability to use a number of different fuel cycles. An active program of analysis and development, to demonstrate the viability of different fuel cycles in CANDU, has been funded by AECL in parallel with the work on the natural uranium cycle. Market forces and advances in technology have obliged us to reassess and refocus some parts of our effort in this area, and significant success has been achieved in integrating all the Canadian efforts in this area. This paper contains a brief summary of some key components of the advanced fuel cycle program. (Author) 4 figs., tab., 18 refs

Next Generation Nuclear Plant Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

P. E. MacDonald

2005-01-01

The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen

Next Generation Nuclear Plant Research and Development Program Plan

Energy Technology Data Exchange (ETDEWEB)

None

2005-01-01

The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: (1) Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission (2) Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented

Development of a computerized nuclear materials control and accounting system for a fuel reprocessing plant

International Nuclear Information System (INIS)

Crawford, J.M.; Ehinger, M.H.; Joseph, C.; Madeen, M.L.

1979-01-01

A computerized nuclear materials control and accounting system (CNMCAS) for a fuel reprocessing plant is being developed by Allied-General Nuclear Services at the Barnwell Nuclear Fuel Plant. Development work includes on-line demonstration of near real-time measurement, measurement control, accounting, and processing monitoring/process surveillance activities during test process runs using natural uranium. A technique for estimating in-process inventory is also being developed. This paper describes development work performed and planned, plus significant design features required to integrate CNMCAS into an advanced safeguards system. 2 refs

Development of nuclear fuel cycle technology

International Nuclear Information System (INIS)

Kawahara, Akira; Sugimoto, Yoshikazu; Shibata, Satoshi; Ikeda, Takashi; Suzuki, Kazumichi; Miki, Atsushi.

1990-01-01

In order to establish the stable supply of nuclear fuel as an important energy source, Hitachi ltd. has advanced the technical development aiming at the heightening of reliability, the increase of capacity, upgrading and the heightening of performance of the facilities related to nuclear fuel cycle. As for fuel reprocessing, Japan Nuclear Fuel Service Ltd. is promoting the construction of a commercial fuel reprocessing plant which is the first in Japan. The verification of the process performance, the ensuring of high reliability accompanying large capacity and the technical development for recovering effective resources from spent fuel are advanced. Moreover, as for uranium enrichment, Laser Enrichment Technology Research Association was founded mainly by electric power companies, and the development of the next generation enrichment technology using laser is promoted. The development of spent fuel reprocessing technology, the development of the basic technology of atomic process laser enrichment and so on are reported. In addition to the above technologies recently developed by Hitachi Ltd., the technology of reducing harm and solidification of radioactive wastes, the molecular process laser enrichment and others are developed. (K.I.)

Dry process fuel performance technology development

International Nuclear Information System (INIS)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

2006-06-01

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

Dry process fuel performance technology development

Energy Technology Data Exchange (ETDEWEB)

Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

2006-06-15

The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options

International Nuclear Information System (INIS)

Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J.; Parma, Edward J.Jr; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

2007-01-01

The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents

Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

Energy Technology Data Exchange (ETDEWEB)

Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

2007-10-01

The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

LSDS Development for Isotopic Fissile Assay in Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-07-01

As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel.

LSDS Development for Isotopic Fissile Assay in Spent Fuel

International Nuclear Information System (INIS)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

2011-01-01

As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel

Enviromental Development Plan: special nuclear materials production

International Nuclear Information System (INIS)

1980-07-01

This Environmental Development Plan includes the process steps and facilities necessary for the production of plutonium and tritium for Government needs and the production of some other radioactive materials that will be used for heat and radiation sources by domestic and international customers. The production reactors and the spent fuel processing plants and their effluents are discussed, but the defense wastes from them are treated in a separate EDP. The scope does not include transportation, decontamination and decommissioning, safeguards and security, or use of the SNM products

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

Energy Technology Data Exchange (ETDEWEB)

Diamond, David

2017-12-29

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allow for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.

Concerning 1991 basic plan for atomic energy development and application (subjected to examination)

International Nuclear Information System (INIS)

1990-01-01

The prime minister developed a draft 1991 Basic Plan for Atomic Energy Development and Application and sent it to the Nuclear Safety Commission for examination. The Commission started the examination at its 14th meeting. The report outlines results of the examination. A Basic Plan is developed each year to promote efforts at atomic energy development and application systematically and efficiently. In particular, it identifies specific activities required to realize the basic policies shown in the Long Term Program for Atomic Energy Development and Application. In the present report, activities required for improving the safety measures in general are described first, with special emphasis placed on the improvement in nuclear safety regulations and promotion of nuclear safety research. Activities required for promoting nuclear power generation are then outlined. It also insists that the nuclear fuel cycle should be established by promoting measures for uranium resources, uranium enrichment, spent fuel enrichment, and radioactive waste disposal. Other required efforts include the development of improved power reactors, implementation of major projects, and development of basic technology. (N.K.)

Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

International Nuclear Information System (INIS)

Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

2015-01-01

This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

Energy Technology Data Exchange (ETDEWEB)

Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-04-01

This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

Operational Readiness Review Implementation Plan for the K Basin Fuel Transfer System

International Nuclear Information System (INIS)

DAVIES, T.H.

2002-01-01

This implementation plan has been prepared to comply with the requirements of U.S. Department of Energy (DOE) Order 425.1A, Startup and Restart of Nuclear Facilities, and DOE-STD-3006-2000, Planning and Conduct of Operational Readiness Reviews (ORR) (DOE 2002). The scope of the ORR is described in the contractor K Basin Fuel Transfer System (FTS) Plan of Action (POA), which was prepared by Spent Nuclear Fuel (SNF) Project line management and approved by the DOE Richland Operations Office (RL) Manager on April 4, 2002 (FH 2002a). While the Project Hanford Management Contractor has been revised to include DOE Order 425.1B, the contractor implementing procedure, ''F-PRO-055, Startup Readiness (Revision 9) has not yet been approved by RL for contractor use. Appendix A provides a crosswalk between the requirements of DOE Order 425.1A and DOE Order 425.1B to show that all requirements of DOE 425.1B are covered by this implementation plan. DOE Order 425.1B indicates that the Secretarial Officer is the Authorization Authority when substantial modifications are made to a Hazard Category 2 nuclear facility. This Authorization Authority has been delegated to the RL Manager by memorandum from Jessie Hill Roberson, dated November 20, 2001 (Roberson 2001). The scope of the ORR is described in the RL Plan of Action, K Basin Fuel Transfer System, prepared by DOE project line management and approved by the RL Manager, the designated approval authority, on September 12, 2002 (Schlender 2002). This implementation plan provides the overall approach and guidelines for performance of the DOE ORR. Appendix B contains the Criteria and Review Approach Documents (CRAD), which define the review objectives and criteria as well as the approach for assessing each objective. ORR results will be published in a final report, as discussed in Section 9.4

Molybdenum-base cermet fuel development

International Nuclear Information System (INIS)

Gurwell, W.E.; Moss, R.W.; Pilger, J.P.; White, G.D.

1987-07-01

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermet. This cermet is to have a high matrix density (≥95%) for high strength and high thermal conductance coupled with a high particle (UN) porosity (∼25%) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous UN microspheres become available. Process development has been conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and vacuum hot press consolidation techniques. This paper summarizes the status of these activities

The development and revision of the Federal Radiological Emergency Response Plan

International Nuclear Information System (INIS)

Gant, K.S.; Adler, M.V.; Wolff, W.F.

1989-01-01

Since 1985, federal agencies have been using the Federal Radiological Emergency Response Plan (FRERP) in exercises and real events. This experience and the development of other emergency response guidance (e.g., National System for Emergency Coordination) are fueling current efforts to review and revise the FRERP to reflect what the agencies have learned since the FRERP was published. Revision efforts are concentrating on clarifying the plan and addressing deficiencies. No major changes are expected in the general structure of the federal response nor should states need to revise their plans because of these modifications. 5 refs

Community development planning

International Nuclear Information System (INIS)

Gray, S.I.

1983-01-01

The focus of this paper will be methods of local community involvement in the community development planning efforts which will be required at the recommended sites. Community development planning will include capital improvement plans, housing plans, zoning changes, business development plans and other community service and fiscal plans required to meet the projected needs of new residents as a result of the repository construction and operation. This paper will present, (1) the need for community planning, (2) methods of responding to community planning needs, (3) current community planning issues to be addressed. 2 references, 1 figure

RERTR program activities related to the development and application of new LEU fuels

International Nuclear Information System (INIS)

Travelli, A.

1983-01-01

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U 3 Si 2 -Al and U 3 Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 3 which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years

Development of Coated Particle Fuel Technology

International Nuclear Information System (INIS)

Cho, Moon Sung; Kim, B. G.; Kim, Y. K.

2009-04-01

UO 2 kernel fabrication technology was developed at the lab sacle(20∼30g-UO 2 /batch). The GSP technique, modified method of sol-gel process, was used in the preparation of spherical ADU gel particle and these particles were converted to UO 3 and UO 2 phases in calcination furnace and sintering furnace respectively. Based on the process variables optimized using simulant kernels in 1-2 inch beds, SiC TRISO-coated particles were fabricated using UO 2 kernel. Power densities of TRISO coated particle fuels and gamma heat of the tubes are calculated as functions of vertical location of the fuel specimen in the irradiation holes by using core physics codes, MCNP and Helios. A finite model was developed for the calculations of temperatures and stresses of the specimen and the irradiation tubes. Dimensions of the test tubes are determined based on the temperatures and stresses as well as the gamma heat generated at the given condition. 9 modules of the COPA code (MECH, FAIL, TEMTR, TEMBL, TEMPEB, FPREL, MPRO, BURN, ABAQ), the MECH, FAIL, TEMTR, TEMBL, TEMPEB, and FPREL were developed. The COPA-FPREL was verified through IAEA CRP-6 accident benchmarking problems. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. The inspection and test plan describing specifications and inspection method of coated particles was developed to confirm the quality standard of coated particles. The quality inspection instructions were developed for the inspection of coated particles by particle size analyzer, density inspection of coating layers by density gradient column, coating thickness inspection by X-ray, and inspection of optical anistropy factor of PyC layer. The quality control system for the TRISO-coated particle fuel was derived based on the status of quality control systems of other countries

Progress and status of the integral fast reactor (IFR) fuel cycle development

International Nuclear Information System (INIS)

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

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

Strategic backdrop analysis for fossil fuel planning. Task 1. Default Case. Report 468-117-07/01

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

This report presents data describing a default case analysis performed using the strategic backdrop analytical framework developed to facilitate fossil fuel planning within the DOE. Target years are 1985 and 2025. Residential, commercial, and industrial energy demands are forecast as well as the impacts of energy technology implementation and market penetration using a set of energy technology assumptions. (DMC)

Strategic backdrop analysis for fossil fuel planning. Task 1. Default Case. Report 468-117-07/02

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

This report presents data describing a default case analysis performed using the strategic backdrop analytical framework developed to facilitate fossil fuel planning within the DOE. Target years are 1985, 2000, and 2025. Residential, commercial, and industrial energy demands and impacts of energy technology implementation and market penetration are forecast using a set of energy technology assumptions. (DMC)

Development of a computer program for the cost analysis of spent fuel management

International Nuclear Information System (INIS)

Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won; Cha, Jeong Hun; Whang, Joo Ho

2009-01-01

So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

Logistic Fuel Processor Development

National Research Council Canada - National Science Library

Salavani, Reza

2004-01-01

... to light gases then steam reform the light gases into hydrogen rich stream. This report documents the efforts in developing a fuel processor capable of providing hydrogen to a 3kW fuel cell stack...

Spent nuclear fuel retrieval system fuel handling development testing. Final report

International Nuclear Information System (INIS)

Jackson, D.R.; Meeuwsen, P.V.

1997-09-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

Test plan for surface and subsurface examinations of K-east and K-west fuel elements

International Nuclear Information System (INIS)

Pitner, A.L.

1997-01-01

The test plan for subsurface examinations on damaged K East and K West Basin fuel elements is presented. The purpose of these examinations is to inspect damaged areas on the fuel elements for the presence of voids, sludge, or broken fuel, and to obtain samples from the damaged areas for subsequent characterization tests

Interim report spent nuclear fuel retrieval system fuel handling development testing

Energy Technology Data Exchange (ETDEWEB)

Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

1997-06-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

Fuel cycle industrialization program prepared by N-Fuel Research Committee, ANRE

Energy Technology Data Exchange (ETDEWEB)

1978-09-01

To meet the new situation resulting from the scaling down of nuclear power development plan in Japan, and the changes due to the new U.S. nuclear non-proliferation policy, the Nuclear Fuel Research Committee of the Agency of Natural Resources and Energy of MITI has prepared the ''Interim Report on the Nuclear Fuel Cycle''. It sets out in precise terms the methods that should be followed for establishing the nuclear fuel cycle in Japan. Major items treated in this report are; uranium ore development, promotion of uranium stockpiling, construction of domestic uranium enrichment plant, promotion of the construction of a nuclear fuel park, Pu utilization and cooperation in international movement for nuclear non-proliferation, and the establishment of measures for radioactive waste management. Discussions are made from technological, economical, and political view points. Also attached are a table of the comprehensive industrialization plan up to the year 2000 and a table of estimated nuclear fuel demand and supply in Japan.

Canadian fuel development program in 1997/98

International Nuclear Information System (INIS)

Lau, J.H.; Kohn, E.; Sejnoha, R.; Cox, D.S.; Macici, N.N.; Steed, R.G.

1997-01-01

This paper describes the CANDU fuel development activities in Canada during 1997 through 1998. The activities include those of the Fuel Technology Program sponsored by the CANDU Owners Group. The goal of the Fuel Technology Program is to maintain and improve the reliability, economics and safety of CANDU fuel in operating reactors. These activities, therefore, concentrate on the present designs of 28-element and 37-element fuel bundles. The Canadian fuel development activities also include those of the Advanced Fuel and Fuel Cycle Technology Program at AECL. These activities concentrate on the development of advanced fuel designs and advanced fuel cycles, which among other advantages, can reduce the capital and fuelling costs, maintain operating margins in aging reactors, improve natural-uranium utilization, and reduce the amount of spent fuel. (author)

H2 Mobilite France - Study for a Fuel Cell Electric Vehicle national deployment plan

International Nuclear Information System (INIS)

2014-01-01

'Mobilite Hydrogene France', a consortium of private and public stakeholders, united by the French Association for Hydrogen and Fuel Cells (AFHYPAC), launched a study in July 2013 to evaluate the potential of Hydrogen and Fuel Cells in French transport. The results provide true perspectives of a deployment plan for the national territory. This French study was implemented in parallel with various initiatives initiated in other European countries (Germany, UK, Denmark, The Netherlands, Sweden..). It was funded by the participants themselves and by the European Union under the HIT (Hydrogen Infrastructure for Transport) framework. On the basis of shared economic data, synchronised deployment scenarios for vehicles and refuelling stations were developed. These showed the environmental, economic and societal benefits of a transition towards Hydrogen Mobility. Hydrogen Mobility appears as a key factor of the Energy Transition. It can make a significant contribution to emission reduction of the transport sector. The established plan of the Consortium proposes an ambitious deployment in France of Hydrogen Fuel Cell Electric Vehicles, together with refuelling stations by 2030. This plan starts today, with as a first step, a progressive deployment of captive fleets. Like in other countries initiatives worldwide, the success of this deployment in France is a function of combined commitments and efforts by private industrial companies and public stakeholders (at both national and regional level). The main challenge now is in the adaptation of the regulatory frameworks and the implementation of funding, within the coordination of the European Union. The consortium is committed to implement The consortium first deployments in French early adopter territories that have embraced already the strategy of a sustainable and de-carbonized mobility

Fuel cell two-wheelers : good market potential in Shanghai and suggestions for development

International Nuclear Information System (INIS)

Li, L.; Yu, Z.; Gong, H.

2005-01-01

A feasibility study on the market potential of 2-wheel fuel cell vehicles in Shanghai, China was presented. The Chinese government has recently invested a considerable amount of money in fuel cell technology research programs. However, 2-wheel fuel cell vehicles are not included in the program's plans. Considering future transportation models and examining market demand, it was anticipated that the 2-wheel fuel cell vehicles have an enormous market potential in China. An outline of market demands was presented. Issues concerning technological availability, hydrogen supply availability and government policies were considered. An economic overview of Shanghai was presented. International trends in fuel cell vehicles were reviewed, as well as issues concerning traffic and environmental impacts. Recommended policies for development were presented. Supply and demand issues were also considered. At present, 2-wheelers are only at the conceptual and demonstration stage. A graduated commercial framework for fuel cell bicycles and scooters was presented, which included initial production and medium scale production phases as well as the development of a hydrogen supply network over a period of 9 years. It was concluded that, as there are more than 9 million bicycles and over 300,000 motorcycles in Shanghai, there is potentially a huge market and a good base the development phases and gradual commercialization of 2-wheel fuel cell vehicles. 9 refs., 4 tabs., 3 figs

Advanced LWR Nuclear Fuel Cladding Development

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Griffith, G.

2012-01-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

Designing an energy planning concept for enhancing the dissemination of renewable energy technologies in developing countries

DEFF Research Database (Denmark)

Lybæk, Rikke; Andersen, Jan; Lund, Søren

2014-01-01

This paper stresses the need for adapting a sustainable energy planning concept, which can support the implementation of renewable energy in developing countries; exemplified by a Vietnamese case. Many developing countries heavily rely on fossil fuel resources and will face energy supply security...... countries, while relevant policies, tools and plans etc. simultaneously are being deployed, enhancing the framework conditions for renewable energy implementation...

Fast breeder fuel element development

International Nuclear Information System (INIS)

Marth, W.; Muehling, G.

1983-08-01

This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

DOE not planning to accept spent nuclear fuel

International Nuclear Information System (INIS)

Anon.

1995-01-01

Samuel K. Skinner, president of Commonwealth Edison Co. (ComEd), said open-quotes The federal government has a clear responsibility to begin accepting spent nuclear fuel in 1988,close quotes citing the Nuclear Waste Policy Act of 1982 before the Senate Energy and Natural Resources Committee. Based in Chicago, ComEd operates 12 nuclear units, making it the nation's largest nuclear utility. open-quotes Since 1983, the consumers who use electricity produced at all nuclear power plants have been paying to fund federal management of spent nuclear fuel. Consumer payments and obligations, with interest, now total more than $10 billion. Electricity consumers have held up their side of the deal. The federal government must do the same,close quotes Skinner added. Skinner represented the Nuclear Energy Institute (NEI) before the committee. NEI is the Washington-based trade association of the nuclear energy industries. For more than 12 years, utility customers have been paying one-tenth of a cent per kWhr to fund a federal spent fuel management program under the Nuclear Waste Policy Act of 1982. Under this act, the federal government assumed responsibility for management of spent fuel from the nation's nuclear power plants. The U.S. Department of Energy (DOE) was assigned to manage the storage and disposal program. DOE committed to begin accepting spent fuel from nuclear power plants by January 31, 1988. DOE has spent almost $5 million studying a site in Nevada, but is about 12 years behind schedule and does not plan to accept spent fuel beginning in 1998. DOE has said a permanent storage site will not be ready until 2010. This poses a major problem for many of the nation's nuclear power plants which supply about 20% of the electricity in the US

IFR fuel cycle--pyroprocess development

International Nuclear Information System (INIS)

Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

1992-01-01

The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

Characterization program management plan for Hanford K basin spent nuclear fuel

International Nuclear Information System (INIS)

TRIMBLE, D.J.

1999-01-01

The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng

Plan for safety case of spent fuel repository at Olkiluoto

International Nuclear Information System (INIS)

Vieno, T.; Ikonen, A.T.K.

2005-02-01

Posiva aims to present the Safety Case supporting the construction license application of the spent fuel repository at Olkiluoto by 2012. An outline and preliminary assessments will be presented in 2009. Interim reporting and an update of the Safety Case plan will be presented in 2006, as required by the authorities. The KBS-3 disposal concept aims at long-term isolation and containment of spent fuel assemblies in durable copper-iron canisters emplaced in a repository to be constructed at a depth between 400 and 600 metres in crystalline bedrock. By 2012, studies on the KBS-3 disposal concept and site investigations at Olkiluoto will have been continued over about thirty years. The construction of an underground rock characterisation facility (called ONKALO) was started in June 2004. The investigations are carried out in close cooperation with the Swedish SKB developing and assessing the same disposal concept at candidate sites, resembling Olkiluoto, at the other side of the Baltic Sea. A safety case is the synthesis of evidence, analyses and arguments that quantify and substantiate the safety, and the level of expert confidence in the safety, of a planned repository. Posiva's Safety Case will be organised in a portfolio including ten main reports, which will be periodically updated according the overall schedule presented in the plan. The Site report describing the present state and past evolution of the Olkiluoto site, as well as the disturbances caused by the construction of ONKALO and the first stage of the repository, forms the geoscientific basis of the Safety Case. The engineering basis is provided by the reports on the Characteristics of spent fuel, Canister design, and Repository design. The Process report containing descriptions and analyses of features, events and processes potentially affecting the disposal system, and the report on the Evolution of site and repository form the scientific basis of the Safety Case. The latter report will describe and

Hanford Site Development Plan

International Nuclear Information System (INIS)

Rinne, C.A.; Curry, R.H.; Hagan, J.W.; Seiler, S.W.; Sommer, D.J.; Yancey, E.F.

1990-01-01

The Hanford Site Development Plan (Site Development Plan) is intended to guide the short- and long-range development and use of the Hanford Site. All acquisition, development, and permanent facility use at the Hanford Site will conform to the approved plan. The Site Development Plan also serves as the base document for all subsequent studies that involve use of facilities at the Site. This revision is an update of a previous plan. The executive summary presents the highlights of the five major topics covered in the Site Development Plan: general site information, existing conditions, planning analysis, Master Plan, and Five-Year Plan. 56 refs., 67 figs., 31 tabs

Hanford Site Development Plan

Energy Technology Data Exchange (ETDEWEB)

Rinne, C.A.; Curry, R.H.; Hagan, J.W.; Seiler, S.W.; Sommer, D.J. (Westinghouse Hanford Co., Richland, WA (USA)); Yancey, E.F. (Pacific Northwest Lab., Richland, WA (USA))

1990-01-01

The Hanford Site Development Plan (Site Development Plan) is intended to guide the short- and long-range development and use of the Hanford Site. All acquisition, development, and permanent facility use at the Hanford Site will conform to the approved plan. The Site Development Plan also serves as the base document for all subsequent studies that involve use of facilities at the Site. This revision is an update of a previous plan. The executive summary presents the highlights of the five major topics covered in the Site Development Plan: general site information, existing conditions, planning analysis, Master Plan, and Five-Year Plan. 56 refs., 67 figs., 31 tabs.

The DUPIC fuel development program in KAERI

Energy Technology Data Exchange (ETDEWEB)

Yang, M S; Park, H S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

French development program on fuel cycle

International Nuclear Information System (INIS)

Viala, M.; Bourgeois, M.

1991-01-01

The need to close the fuel cycle of fast reactors makes the development of the cycle installations (fuel fabrication, irradiated assembly conditioning before reprocessing, reprocessing and waste management) especially independent with the development of the reactor. French experience with the integrated cycle over a period of about 25 years, the tonnage of fuels fabricated (more than 100 t of mixed oxides) for the Rapsodie, Phoenix and SuperPhoenix reactors, and the tonnage of reprocessed fuel (nearly 30 t of plutonium fuel) demonstrate the control of the cycle operations. The capacities of the cycle installations in existence and under construction are largely adequate for presents needs, even including a new European EFR reactor. They include the Cadarache fuel fabrication complex, the La Hague UP2-800 reprocessing plant, and the Marcoule pilot facility. Short- and medium-term R and D programs are connected with fuel developments, with the primary objective of very high burnups. For the longer term and for a specific plant to reprocess fast reactor fuels, the programs could concern new fabrication and reprocessing systems and the study of the consequences of the reduction in fuel out-of-core time

Development of boats propelled by fuel cells; Developpement d'embarcations mues par piles a combustible

Energy Technology Data Exchange (ETDEWEB)

Affolter, J.F.; Mariotti, P. [Ecole d' Ingenieurs du Canton de Vaud, HES-SO, Yverdon les Bains (Switzerland)

2000-07-01

An important European demand for propulsion systems lead us to think that the number of electric ships is growing rapidly. A lot of lakes in Europe are not allowed to the thermal navigation. The aim is to offer a maximum of comfort, reliability, silence, lack of pollution and respect to environment. One of the answers to this expectation is the use of the electric propulsion and a generation of power with fuel cells. After having fabricated the 'Hydroxy100', a single-seater boat propelled by a 100 W fuel cell, and a two-seater boat designed for a 300 W fuel cell, a new plan for a six-seater pleasure boat is being studied. These pilot plans are carried out with the aim to favour the development of this technology. (O.M.)

Oxide fuel fabrication technology development of the FaCT project (5). Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

International Nuclear Information System (INIS)

Ohtsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

2011-01-01

This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production, both of which are key points for the commercialized use of ODS steels as long-life fuel cladding tubes. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed, where the handling of elemental powder is prohibited in the process. (author)

Unified fuel elements development for research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1998-01-01

Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

History of fast reactor fuel development

Energy Technology Data Exchange (ETDEWEB)

Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

1993-09-01

The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

Metallic fuel design development

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual design technology on metallic fuel

Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

International Nuclear Information System (INIS)

Freedman, J.M.; Humphreys, J.R.

1975-10-01

A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

Nuclear fuel fabrication - developing indigenous capability

International Nuclear Information System (INIS)

Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K.

1997-01-01

Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO 2 powder production are also briefly indicated. (author)

Spent nuclear fuels project: FY 1995 multi-year program plan, WBS {number_sign}1.4

Energy Technology Data Exchange (ETDEWEB)

Denning, J.L.

1994-09-01

The mission of the Spent Nuclear Fuel (SNF) program is to safely, reliably, and efficiently manage, condition, transport, and store Department of Energy (DOE)-owned SNF, so that it meets acceptance criteria for disposal in a permanent repository. The Hanford Site Spent Nuclear Fuel strategic plan for accomplishing the project mission is: Establish near-term safe storage in the 105-K Basins; Complete national Environmental Policy Act (NEPA) process to obtain a decision on how and where spent nuclear fuel will be managed on the site; Define and establish alternative interim storage on site or transport off site to support implementation of the NEPA decision; and Define and establish a waste package qualified for final disposition. This report contains descriptions of the following: Work Breakdown Structure; WBS Dictionary; Responsibility Assignment Matrix; Program Logic Diagrams; Program Master Baseline Schedule; Program Performance Baseline Schedule; Milestone List; Milestone Description Sheets; Cost Baseline Summary by Year; Basis of Estimate; Waste Type Data; Planned Staffing; and Fiscal Year Work Plan.

Spent nuclear fuels project: FY 1995 multi-year program plan, WBS number-sign 1.4

International Nuclear Information System (INIS)

Denning, J.L.

1994-09-01

The mission of the Spent Nuclear Fuel (SNF) program is to safely, reliably, and efficiently manage, condition, transport, and store Department of Energy (DOE)-owned SNF, so that it meets acceptance criteria for disposal in a permanent repository. The Hanford Site Spent Nuclear Fuel strategic plan for accomplishing the project mission is: Establish near-term safe storage in the 105-K Basins; Complete national Environmental Policy Act (NEPA) process to obtain a decision on how and where spent nuclear fuel will be managed on the site; Define and establish alternative interim storage on site or transport off site to support implementation of the NEPA decision; and Define and establish a waste package qualified for final disposition. This report contains descriptions of the following: Work Breakdown Structure; WBS Dictionary; Responsibility Assignment Matrix; Program Logic Diagrams; Program Master Baseline Schedule; Program Performance Baseline Schedule; Milestone List; Milestone Description Sheets; Cost Baseline Summary by Year; Basis of Estimate; Waste Type Data; Planned Staffing; and Fiscal Year Work Plan

Battery and Fuel Cell Development Goals for the Lunar Surface and Lander

Science.gov (United States)

Mercer, Carolyn R.

2008-01-01

NASA is planning a return to the moon and requires advances in energy storage technology for its planned lunar lander and lunar outpost. This presentation describes NASA s overall mission goals and technical goals for batteries and fuel cells to support the mission. Goals are given for secondary batteries for the lander s ascent stage and suits for extravehicular activity on the lunar surface, and for fuel cells for the lander s descent stage and regenerative fuel cells for outpost power. An overall approach to meeting these goals is also presented.

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

International Nuclear Information System (INIS)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

1999-04-01

A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

Development of CANFLEX fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan

1991-12-01

This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle(so-called CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactors for 1996 and 1997, and consequently will be used in the existing and future reactors in Korea. The research activities during this year include the basic design of CANFLEX fuel with slightly enriched uranium(CANFLEX-SEU), with emphasis on the extension of fuel operation limit. Based on this basic design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel. (Author)

A decision analysis framework to support long-term planning for nuclear fuel cycle technology research, development, demonstration and deployment

International Nuclear Information System (INIS)

Sowder, A.G.; Machiels, A.J.; Dykes, A.A.; Johnson, D.H.

2013-01-01

To address challenges and gaps in nuclear fuel cycle option assessment and to support research, develop and demonstration programs oriented toward commercial deployment, EPRI (Electric Power Research Institute) is seeking to develop and maintain an independent analysis and assessment capability by building a suite of assessment tools based on a platform of software, simplified relationships, and explicit decision-making and evaluation guidelines. As a demonstration of the decision-support framework, EPRI examines a relatively near-term fuel cycle option, i.e., use of reactor-grade mixed-oxide fuel (MOX) in U.S. light water reactors. The results appear as a list of significant concerns (like cooling of spent fuels, criticality risk...) that have to be taken into account for the final decision

Spent fuel management in the Republic of Korea: Current status and plans

International Nuclear Information System (INIS)

Sang Doug Park

1998-01-01

Korea has selected nuclear energy as the major source for the electric power generation due to the insufficiency of energy resources in Korea. in compliance with the policy, Korea Electric Power Corporation (KEPCO) has expanded the nuclear power programme and faced the significant arisings of spent fuel. The interim At Reactor(AR) storage pools have very limited capacities and temporary expansion of this capacity has been taken such as re-racking and dry storage construction. There was a plan, to construct a centralized spent fuel storage facility, which was postponed officially by the government. Under the current situation, it is hard to establish the long-term spent fuel management strategy. 'Wait and See' is no more applicable to Korea. because of storage shortage. Within R and D, dry storage construction and DUPIC fuel cycle are being considered. In this paper, the spent fuel management programme of Korea is briefly reviewed. (author)

Development of CH{sub 3}OH fueled PEMFC power plants for hybrid transit buses

Energy Technology Data Exchange (ETDEWEB)

Baumert, R; Cooper, R; Feasey, G [DBB Fuel Cell Engines Corp., Poway, CA (United States)

1999-12-31

An overview of the methanol fuel cell power system was provided, identifying improved efficiency and reduced emissions as the principal advantages. Four critical tasks regarding on-board fuel processing were described: (1) efficient methanol conversion (steam reforming), (2) effective reformate purification (selective catalytic oxidation), (3) optimized heat integration, and (4) rapid response to transients. A description of a 100 kW PEM fuel cell bus engine package was also presented. As far as a development time table is concerned, the DBB Fuel Cell Engines Corp. of Poway California has completed two methanol fueled PEMFC power plants, fabrication of the initial 100 kW PEMFC engine is in progress and scheduled for delivery by 1998. The two methanol fueled commercial products which are in the planning stages are the 100 and 200 kW class FCPS for hybrid and non-hybrid buses and other applications. tabs., figs.

Development of FR fuel cycle in japan (1) development scope of fuel cycle technology

International Nuclear Information System (INIS)

Nakamura, H.; Funasaka, H.; Namekawa, T.

2008-01-01

A fast reactor (FR) cycle has a potential to realize a sustainable energy supply system that is harmonized with environment by fully recycling both uranium (U) and transuranium (TRU) elements. In Japan, a Feasibility Study on Commercialized FR Cycle Systems (FS) was launched in July 1999, and through two different study phases, a final report was presented in 2006. As a result of FS, a combined system of sodium-cooled FR with mixed-oxide (MOX) fuel, advanced aqueous reprocessing and simplified pelletizing fuel fabrication was considered to be most promising for commercialization. The advanced aqueous reprocessing system, which is called the New Extraction system for TRU recovery (NEXT), consists of a U crystallization process for the bulk of U recovery, a simplified solvent extraction process for residual U, plutonium (Pu) and neptunium (Np) without Pu partitioning and purification, and a process for recovering americium (Am) and curium (Cm) from the raffinate. The ratio of Pu/U concentration in the mother solution after crystallization is adequate for MOX fuel fabrication, and thus complicated powder mixing processes for adjusting Pu content in MOX fuel can be eliminated in the subsequent simplified fuel fabrication system. In this system, lubricant-mixing process can also be eliminated by adopting the advanced technology in which lubricant is coated on the inner surface of a die before fuel powder supply. Such a simplification could help us overcoming the difficulty to treat MA bearing fuel powders in a hot cell. Ministry of Education, Culture, Sports, Science and Technology (MEXT) reviewed these results of FS in 2006 and identified the most promising FR cycle concept proposed in the FS phase II study as a mainline choice for commercialization. According to such a governmental assessment, R and D activities of FR cycle systems were decided to be concentrated mainly to the innovative technology development for the mainline concept. The stage of R and D project was

Plans for the development of the IFR [Integral Fast Reactor] fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.

1986-01-01

The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium

Development of high burnup nuclear fuel technology

International Nuclear Information System (INIS)

Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

1987-04-01

The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

Stochastic Programming for Fuel Supply Planning of Combined Heat and Power Plants

DEFF Research Database (Denmark)

Guericke, Daniela; Blanco, Ignacio; Morales González, Juan Miguel

The consumption of biomass to produce power and heat has increased due to the carbon neutral policies. Combined heat and power (CHP) plants often combine biomass with other fuels, e.g., natural gas. The negotiation process for supply contracts involves many uncertainties due to the long planning...... horizon. The demand for biomass is uncertain, and heat demand and electricity prices vary during the planning period. We propose a method using stochastic optimization to support the biomass and natural gas supply planning for CHP plants including short-term decisions for optimal market participation....

Romanian nuclear fuel cycle development

International Nuclear Information System (INIS)

Rapeanu, S.N.; Comsa, Olivia

1998-01-01

Romanian decision to introduce nuclear power was based on the evaluation of electricity demand and supply as well as a domestic resources assessment. The option was the introduction of CANDU-PHWR through a license agreement with AECL Canada. The major factors in this choice have been the need of diversifying the energy resources, the improvement the national industry and the independence of foreign suppliers. Romanian Nuclear Power Program envisaged a large national participation in Cernavoda NPP completion, in the development of nuclear fuel cycle facilities and horizontal industry, in R and D and human resources. As consequence, important support was being given to development of industries involved in Nuclear Fuel Cycle and manufacturing of equipment and nuclear materials based on technology transfer, implementation of advanced design execution standards, QA procedures and current nuclear safety requirements at international level. Unit 1 of the first Romanian nuclear power plant, Cernavoda NPP with a final profile 5x700 Mw e, is now in operation and its production represents 10% of all national electricity production. There were also developed all stages of FRONT END of Nuclear Fuel Cycle as well as programs for spent fuel and waste management. Industrial facilities for uranian production, U 3 O 8 concentrate, UO 2 powder and CANDU fuel bundles, as well as heavy water plant, supply the required fuel and heavy water for Cernavoda NPP. The paper presents the Romanian activities in Nuclear Fuel Cycle and waste management fields. (authors)

Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

Energy Technology Data Exchange (ETDEWEB)

Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

2014-05-15

700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

LOFT advanced fuel rod instrumentation development

International Nuclear Information System (INIS)

Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

1978-01-01

Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

LANDFIRE - A national vegetation/fuels data base for use in fuels treatment, restoration, and suppression planning

Science.gov (United States)

Kevin C. Ryan; Tonja S. Opperman

2013-01-01

LANDFIRE is the working name given to the Landscape Fire and Resource Management Planning Tools Project (http://www.landfire.gov). The project was initiated in response to mega-fires and the need for managers to have consistent, wall-to-wall (i.e., all wildlands regardless of agency/ownership), geospatial data, on vegetation, fuels, and terrain to support use of fire...

Integrating fire behavior models and geospatial analysis for wildland fire risk assessment and fuel management planning

Science.gov (United States)

Alan A. Ager; Nicole M. Vaillant; Mark A. Finney

2011-01-01

Wildland fire risk assessment and fuel management planning on federal lands in the US are complex problems that require state-of-the-art fire behavior modeling and intensive geospatial analyses. Fuel management is a particularly complicated process where the benefits and potential impacts of fuel treatments must be demonstrated in the context of land management goals...

Fuel cell two wheelers: Good market potential in Shanghai and the suggestions for development. Paper no. IGEC-1-145

International Nuclear Information System (INIS)

Li, L.; Yu, Z.; Gong, H.

2005-01-01

This paper presents the feasibility and the market potential for developing the Fuel Cell two-wheelers in Asia. Shanghai is taken as one case for this analysis. Based on the study of the transportation modal, the constitute of the vehicles both motored and non-motored, the future planning for the transportation and the market potential of two-wheelers requirement and the quick development market of Battery Bicycles in Shanghai, the potential market of Fuel Cell Two-wheelers in Shanghai is predicated. The predominance in the research and development for Fuel Cell stacks, the storage for the hydrogen and the local supply ability of hydrogen of Shanghai are also introduced. The Shanghai's Fuel Cell Market potential is also presented based on the population of motorcycles at last. The suggestions for the further development of Fuel Cell two-wheelers are proposed. (author)

105-N Fuel Storage Basin dewatering conceptual plan

International Nuclear Information System (INIS)

Schilperoort, D.L.

1996-11-01

This conceptual plan discusses the processes that will be used for draining and disposing of water from the 105-N Fuel Storage Basin (N Basin), and includes a description of the activities to control surface contamination and potential high dose rates encountered during dewatering. The 105-N Fuel Storage Basin is located in the 100-N Area of the Hanford Site in Richland, Washington. The processes for water disposal include water filtration, water sampling and analysis, tanker loading and unloading, surface decontamination and sealing, and clean out and disposal of residual debris and sediments during final pumping to remove the N Basin water. Water disposal is critical for the deactivation of N Reactor. A Memorandum of Understanding (MOU) between the US Department of Energy (DOE) Environmental Restoration (ER) Program and DOE Waste Management (WM) Program establishes the 200 East Effluent Treatment Facility (ETF) as the final treatment and disposal site for N Basin water and identifies pre-treatment requirements. This MOU states that water delivery will be completed no later than October 31, 1996, and will require a revision due to the current de-watering schedule date. The current MOU requires four micron filtration prior to shipment to ETF. The MOU revision for delivery date extension seeks to have the filtration limit increased to five microns, which eliminates the need for a second filter system and simplifies dewatering. For the purposes of this plan, it will be assumed that five micron filtration will be used

Metallic fuel design development

Energy Technology Data Exchange (ETDEWEB)

Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

1999-04-01

This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

Plutonium Focus Area research and development plan. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-11-01

The Department of Energy (DOE) committed to a research and development program to support the technology needs for converting and stabilizing its nuclear materials for safe storage. The R and D Plan addresses five of the six material categories from the 94-1 Implementation Plan: plutonium (Pu) solutions, plutonium metals and oxides, plutonium residues, highly enriched uranium, and special isotopes. R and D efforts related to spent nuclear fuel (SNF) stabilization were specifically excluded from this plan. This updated plan has narrowed the focus to more effectively target specific problem areas by incorporating results form trade studies. Specifically, the trade studies involved salt; ash; sand, slag, and crucible (SS and C); combustibles; and scrub alloy. The plan anticipates possible disposition paths for nuclear materials and identifies resulting research requirements. These requirements may change as disposition paths become more certain. Thus, this plan represents a snapshot of the current progress and will continue to be updated on a regular basis. The paper discusses progress in safeguards and security, plutonium stabilization, special isotopes stabilization, highly-enriched uranium stabilization--MSRE remediation project, storage technologies, engineered systems, core technology, and proposed DOE/Russian technology exchange projects.

Plutonium Focus Area research and development plan. Revision 1

International Nuclear Information System (INIS)

1996-11-01

The Department of Energy (DOE) committed to a research and development program to support the technology needs for converting and stabilizing its nuclear materials for safe storage. The R and D Plan addresses five of the six material categories from the 94-1 Implementation Plan: plutonium (Pu) solutions, plutonium metals and oxides, plutonium residues, highly enriched uranium, and special isotopes. R and D efforts related to spent nuclear fuel (SNF) stabilization were specifically excluded from this plan. This updated plan has narrowed the focus to more effectively target specific problem areas by incorporating results form trade studies. Specifically, the trade studies involved salt; ash; sand, slag, and crucible (SS and C); combustibles; and scrub alloy. The plan anticipates possible disposition paths for nuclear materials and identifies resulting research requirements. These requirements may change as disposition paths become more certain. Thus, this plan represents a snapshot of the current progress and will continue to be updated on a regular basis. The paper discusses progress in safeguards and security, plutonium stabilization, special isotopes stabilization, highly-enriched uranium stabilization--MSRE remediation project, storage technologies, engineered systems, core technology, and proposed DOE/Russian technology exchange projects

Work plan: transient release from LMFBR fuel

International Nuclear Information System (INIS)

Kress, T.S.; Parker, G.W.; Fontana, M.H.

1975-09-01

The proposed LMFBR Transient Release Program at ORNL is designed to investigate, by means of ex-reactor experiments and analytical modeling, the release and transport of fuel, fission products, and transuranic elements from fast reactor cores in the event of certain hypothetical accidents. It is desired to experimentally produce energy depositions that are characteristic of severe hypothetical reactor transients by the application of direct electrical current to mixed-oxide fuels under sodium. The experimental program includes tests with and without sodium, investigations of alternative methods of generating fuel and sodium aerosols, the use of UO 2 as a fuel simulant, additions of tracers as fission product simulants, effects of radiation, and under-water and under-sodium efforts to study the behavior of the vapor bubble itself. Analytical modeling will accompany all phases of the program, and the data will be correlated with models developed. 21 references. (auth)

Development of fuel and energy storage technologies

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

Development of fuel cell power plants is intended of high-efficiency power generation using such fuels with less air pollution as natural gas, methanol and coal gas. The closest to commercialization is phosphoric acid fuel cells, and the high in efficiency and rich in fuel diversity is molten carbonate fuel cells. The development is intended to cover a wide scope from solid electrolyte fuel cells to solid polymer electrolyte fuel cells. For new battery power storage systems, development is focused on discrete battery energy storage technologies of fixed type and mobile type (such as electric vehicles). The ceramic gas turbine technology development is purposed for improving thermal efficiency and reducing pollutants. Small-scale gas turbines for cogeneration will also be developed. Development of superconduction power application technologies is intended to serve for efficient and stable power supply by dealing with capacity increase and increase in power distribution distance due to increase in power demand. In the operations to improve the spread and general promotion systems for electric vehicles, load leveling is expected by utilizing and storing nighttime electric power. Descriptions are given also on economical city systems which utilize wide-area energy. 30 figs., 7 tabs.

Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

International Nuclear Information System (INIS)

Wachs, G.W.

1997-11-01

The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

The US department of energy's research and development plans for the use of nuclear energy for hydrogen production

International Nuclear Information System (INIS)

Henderson, A.D.; Pickard, P.S.; Park, C.V.; Kotek, J.F.

2004-01-01

The potential of hydrogen as a transportation fuel and for stationary power applications has generated significant interest in the United States. President George W. Bush has set the transition to a 'hydrogen economy' as one of the Administration's highest priorities. A key element of an environmentally-conscious transition to hydrogen is the development of hydrogen production technologies that do not emit greenhouse gases or other air pollutants. The Administration is investing in the development of several technologies, including hydrogen production through the use of renewable fuels, fossil fuels with carbon sequestration, and nuclear energy. The US Department of Energy's Office of Nuclear Energy, Science and Technology initiated the Nuclear Hydrogen Initiative to develop hydrogen production cycles that use nuclear energy. The Nuclear Hydrogen Initiative has completed a Nuclear Hydrogen R and D Plan to identify candidate technologies, assess their viability, and define the R and D required to enable the demonstration of nuclear hydrogen production by 2016. This paper gives a brief overview of the Nuclear Hydrogen Initiative, describes the purposes of the Nuclear Hydrogen R and D Plan, explains the methodology followed to prepared the plan, presents the results, and discusses the path forward for the US programme to develop technologies which use nuclear energy to produce hydrogen. (author)

Development of low enrichment technologies for high density fuels and for isotope production targets

International Nuclear Information System (INIS)

Taboada, Horacio; Gonzalez, Alfredo G.

2005-01-01

Since more than twenty years ago, CNEA has carried out RERTR activities. Main goals are to convert the RA 6 reactor core from HEU to LEU, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, and to optimize techniques to recover U from silicide scrap samples. The future plans include: 1) Completion the RA 6 reactor conversion to LEU; 2) Qualification by irradiation of the promising solutions found for the high density fuels; 3) Irradiation of mini plates and full scale fuel assemblies at the RA 3 reactor and at higher flux and temperature reactors; 4) Optimization of LEU target and radiochemical techniques for radioisotope production. (author) [es

Japan's fuel recycling policy

International Nuclear Information System (INIS)

Anon.

1991-01-01

The Atomic Energy Commission (AEC) has formulated Japanese nuclear fuel recycling plan for the next 20 years, based on the idea that the supply and demand of plutonium should be balanced mainly through the utilization of plutonium for LWRs. The plan was approved by AEC, and is to be incorporated in the 'Long term program for development and utilization of nuclear energy' up for revision next year. The report on 'Nuclear fuel recycling in Japan' by the committee is characterized by Japanese nuclear fuel recycling plan and the supply-demand situation for plutonium, the principle of the possession of plutonium not more than the demand in conformity with nuclear nonproliferation attitude, and the establishment of a domestic fabrication system of uranium-plutonium mixed oxide fuel. The total plutonium supply up to 2010 is estimated to be about 85 t, on the other hand, the demand will be 80-90 t. The treatment of plutonium is the key to the recycling and utilization of nuclear fuel. By around 2000, the private sector will commercialize the fabrication of the MOX fuel for LWRs at the annual rate of about 100 t. Commitment to nuclear nonproliferation, future nuclear fuel recycling program in Japan, MOX fuel fabrication system in Japan and so on are reported. (K.I.)

The low-enrichment fuel development program

International Nuclear Information System (INIS)

Stahl, D.

1993-01-01

In the 1950s and 1960s, low-power research reactors were built around the world utilized MTR-type fuel elements containing 20% enriched uranium. However, the demand for higher specific power created a need for greater uranium-235 concentrations. Early difficulties in increasing uranium content led to the substitution of highly enriched uranium in place of the 20% enriched fuel previously utilized. The highly enriched material also yielded other benefits including longer core residence time, higher specific reactivity, and somewhat lower cost. Highly enriched material then became readily available and was used for high-power reactors as well as in low-power reactors where 20% enriched material would have sufficed. The trend toward higher and higher specific power also led to the development of the dispersion-type fuels which utilized highly enriched uranium at a concentration of about 40 wt%. In the 1970's, however, concerns were raised about the proliferation resistance of fuels and fuel cycles. As a consequence, the U.S. Department of State has recently prohibited the foreign shipment of highly enriched material, except where prior contractual obligation or special merit exists. This will impact on the availability and utilization of highly enriched uranium for research and test reactor fuel. It has also stimulated development programs on fuels with higher uranium content which would allow the use of uranium of lower enrichment. The purpose of this report is to briefly describe the overall fuel-development program which is coordinated by Argonne National Laboratory for the Department of Energy, and to indicate the current and potential uranium loadings. Other reports will address the individual fuel-development activities in greater detail

Fuel Cell Development and Test Laboratory | Energy Systems Integration

Science.gov (United States)

Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

Brief summary of water reactor fuel activities in China

Energy Technology Data Exchange (ETDEWEB)

Zhongyue, Zhang [China Inst. of Atomic Energy (CIAE), Beijing (China)

1997-12-01

The presentation briefly reviews the water reactor fuel activities in China describing: nuclear power development program and growth forecast; fuel performance;fuel performance code improvement; research and development plans. 1 ref., 3 figs, 2 tabs.

Advanced coal-fueled industrial cogeneration gas turbine system -- combustion development

Energy Technology Data Exchange (ETDEWEB)

LeCren, R.T.

1994-06-01

This topical report summarizes the combustor development work accomplished under the subject contract. The objective was to develop a combustion system for the Solar 4MW Type H Centaur gas turbine generator set which was to be used to demonstrate the economic, technical and environmental feasibility of a direct coal-fueled gas turbine in a 100 hour proof-of-concept test. This program started with a design configuration derived during the CSC program. The design went through the following evolution: CSC design which had some known shortcomings, redesigned CSC now designated as the Two Stage Slagging Combustor (TSSC), improved TSSC with the PRIS evaluated in the IBSTF, and full scale design. Supporting and complimentary activities included computer modelling, flow visualization, slag removal, SO{sub x} removal, fuel injector development and fuel properties evaluation. Three combustor rigs were utilized: the TSSC, the IBSTF and the full scale rig at Peoria. The TSSC rig, which was 1/10th scale of the proposed system, consisted of a primary and secondary zone and was used to develop the primary zone performance and to evaluate SO{sub x} and slag removal and fuel properties variations. The IBSTF rig which included all the components of the proposed system was also 1/10th scale except for the particulate removal system which was about 1/30th scale. This rig was used to verify combustor performance data obtained on the TSSC and to develop the PRIS and the particulate removal system. The full scale rig initially included the primary and secondary zones and was later modified to incorporate the PRIS. The purpose of the full scale testing was to verify the scale up calculations and to provide a combustion system for the proof-of-concept engine test that was initially planned in the program.

GCRA review and appraisal of fuel material development programs

International Nuclear Information System (INIS)

1980-09-01

The Fuel material Development Program has as its principal objective and responsibility the development of a fuel that is both economical and licensable and that, at the same time, will fulfill the required performance criteria. To accomplish this, the program is broken down into the following major fuel development task areas: development of the experimental and analytical data base for selecting, qualifying, and verifying the reference fuel design; providing the data base and developing models for evaluating fuel performance under upset and accident conditions; and developing and justifying fuel fabrication specifications which are consistent with the overall fuel performance criteria and with the fuel fabrication process capabilities

Nuclear Fuel Design Technology Development for the Future Fuel

Energy Technology Data Exchange (ETDEWEB)

Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

2006-04-15

The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

Development of Nuclear Fuel Remote Fabrication Technology

International Nuclear Information System (INIS)

Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

2005-04-01

The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

Developments in fuel manufacturing

International Nuclear Information System (INIS)

Williams, T.

1997-01-01

BNFL has a long tradition of willingness to embrace technological challenge and a dedication to quality. This paper describes advances in the overall manufacturing philosophy at BNFL's Fuel Business Group and then covers how some new technologies are currently being employed in BNFL Fuel Business Group's flagship oxide complex (OFC), which is currently in its final stages of commissioning. This plant represents a total investment of some Pound 200 million. This paper also describes how these technologies are also being deployed in BNFL's MOX plant now being built at Sellafield and, finally, covers some new processes being developed for advanced fuel manufacture. (author)

Subchannel analysis code development for CANDU fuel channel

International Nuclear Information System (INIS)

Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

1998-07-01

Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

Hanford 300 Area Development Plan

International Nuclear Information System (INIS)

Daly, K.S.; Seiler, S.W.; Hail, J.C.

1991-09-01

The purpose of the Hanford 300 Area Development Plan (Development Plan) is to guide the physical development of the 300 Area in accordance with US Department of Energy (DOE) Order 4320.1B (DOE 1991b) by performing the following: (1) Establishing a land use plan, setting land use categories that meet the needs of existing and proposed activities; (2) Coordinating existing, 5-yr, and long-range development plans and guiding growth in accordance with those plans; (3) Establishing development guidelines to encourage cost-effective development and minimize conflicts between adjacent activities; (4) Identifying site development issues that need further analysis; Integrating program plans with development plans to ensure a logical progression of development; and, (6) Integrating DOE plans with local agency plans (i.e., city, country, state, and Tri-Cities Science and Technology Park plans)

Globalization of the nuclear fuel cycle impact of developments on fuel management

Energy Technology Data Exchange (ETDEWEB)

Van Den Durpel, L.; Bertel, E. [OCDE-NEA, Nuclear Development Div., 92 - Issy-les-Moulineaux (France)

1999-07-01

Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the de-regulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to compete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economical perspective including environmental and social considerations. (authors)

Globalisation of the nuclear fuel cycle - impact of developments on fuel management

International Nuclear Information System (INIS)

Durpel, L. van den; Bertel, E.

2000-01-01

Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the deregulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to complete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according to the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economic perspective including environmental and social considerations. (orig.) [de

Training development in Juzbado's Fuel Cycle Facility

International Nuclear Information System (INIS)

Perez, A.; Cunado, E.; Ortiz, D.

2003-01-01

In Juzbado's fuel cycle facility, because of the special activities developed, training is a very important issues. Training has been evolved, due to changes on the standards applicable each moment, and also due to the technological resources available. Both aspects have resulted in an evolution of the documents referred to training, such as training programs procedures, Radiation Protection Manual as well as the teaching methods. In the report we are going to present, we will show more precisely the changes that take place, referring to the different training methods used, present training sanitations, and the improvements already planned in training subjects as well as tools used, accomplishing with the legislation and improving in our effort of a better assimilation of the necessary knowledge. (Author)

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)

Action plan for coordinated deployment of hydrogen fuel cell vehicles and hydrogen infrastructure

International Nuclear Information System (INIS)

Elrick, W.

2009-01-01

This paper discussed a program designed to provide hydrogen vehicles and accessible hydrogen stations for a pre-commercial hydrogen economy in California. The rollout will coordinate the placement of stations in areas that meet the needs of drivers in order to ensure the transition to a competitive marketplace. An action plan has been developed that focuses on the following 3 specific steps: (1) the validation of early passenger vehicle markets, (2) expanded transit bus use, and (2) the establishment of regulations and standards. Specific tasks related to the steps were discussed, as well as potential barriers to the development of a hydrogen infrastructure in California. Methods of ensuring coordinated actions with the fuel cell and hydrogen industries were also reviewed

Development of spent fuel dry storage technology

International Nuclear Information System (INIS)

Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

2000-01-01

The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

Canadian CANDU fuel development program and recent fuel operating experience

International Nuclear Information System (INIS)

Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

1999-01-01

This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

Development of Test Methods for Structural Components of a Dual Cooled Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu; Yoon, Kyung Ho; Kim, Jae Yong

2009-01-15

The most unique feature of a dual-cooled fuel is that the outer diameter of a fuel rod is considerably increased due to an internal coolant passage additionally formed inside the rod. This increases the fuel rod's weight and decreases the gap between the fuel rods. Change of the weight and gap causes the shape and the performance of a fuel rod support structure to be necessarily altered. It also alters the flow-induced vibration (FIV) as well as the fretting wear characteristics of a fuel rod. These are directly related with the integrity of the rod so that they should be investigated in the design stage. Finite element analysis and semi-empirical formulae can be used to roughly investigate the support performance and FIV of a fuel rod, respectively. However, the fretting wear characteristic can be investigated only through an experiment. The support performance and FIV need experiment as well to obtain the characteristics more accurately. Therefore, experimental investigation of those has been included in the present project scope. In the second year, it has been planned to establish the experimental devices and technologies to accommodate the altered dimensions and feature of a dual-cooled fuel. As a result, devices to obtain the characteristics of fuel rod supports and holddown springs were developed. As for the FIV and fretting wear characteristics, the existing facilities were modified. Experimental procedures were also re-established this year.

Ongoing and planned fuel safety research in NEA member states - Compiled from SEGFSM Members' Contributions

International Nuclear Information System (INIS)

2002-10-01

This report is in response to an action placed on SEGFSM members to compile ongoing and planned fuel safety research in NEA member states with the aim of providing CSNI an overview on related R and D international programmes and projects, along with the identification of current and future needs and priorities. A questionnaire was distributed to SEGFSM members on 18 October 2000, requesting them to identify fuel safety research programmes and to provide information on achievements and future plans. The questionnaire required respondents to provide information on the ongoing R and D programmes under the following headings: Title; Research Laboratory/Sponsor(s); Objectives/Goals; Status of Work; Brief description/presentation of the main results achieved; Future plans; References. Replies were received from organizations in the following countries: Belgium, Canada, Czech Republic, France, Germany, Hungary, Japan, Korea, Norway (Halden Reactor Project), Sweden, Switzerland, United Kingdom, USA. The report is based on the information provided in the replies received, as a consequence it cannot be viewed as comprehensive; programmes may well be in progress in addition to those detailed here. It is also possible that the detailed results of some programmes may remain proprietary and therefore not available in the short term. The report is organized in topic sections relating to: fuel and clad studies, integral fuel rod tests and PIE, LOCA and RIA studies including whole rods and bundles as well as single effects studies of fuel and cladding, code development for both steady state and transient fuel behaviour, thermal hydraulics, reactor physics codes and finally severe accident studies. The main issues for the current generation of reactors are those of high burn-up performance in normal operations, LOCA and RIA conditions and the main goal for the industry is to consolidate the safety issues to bring all countries up to a licensed discharge burn-up of ∼60 MWd/kg in

Development of fuel cell systems for aircraft applications based on synthetic fuels

Energy Technology Data Exchange (ETDEWEB)

Pasel, J.; Samsun, R.C.; Doell, C.; Peters, R.; Stolten, D. [Forschungszentrum Juelich GmbH (Germany)

2010-07-01

At present, in the aviation sector considerable scientific project work deals with the development of fuel cell systems based on synthetic fuels to be integrated in future aircraft. The benefits of fuel cell systems in aircraft are various. They offer the possibility to simplify the aircraft layout. Important systems, i.e. the gas turbine powered auxiliary power unit (APU) for electricity supply, the fuel tank inserting system and the water tank, can be substituted by one single system, the fuel cell system. Additionally, the energy demand for ice protection can be covered assisted by fuel cell systems. These measures reduce the consumption of jet fuel, increase aircraft efficiency and allow the operation at low emissions. Additionally, the costs for aircraft related investments, for aircraft maintenance and operation can be reduced. On the background of regular discussions about environmental concerns (global warming) of kerosene Jet A-1 and its availability, which might be restricted in a few years, the aircraft industry is keen to employ synthetic, sulfur-free fuels such as Fischer-Tropsch fuels. These comprise Bio-To-Liquid and Gas-To-Liquid fuels. Within this field of research the Institute of Energy Research (IEF-3) in Juelich develops complete and compact fuel cell systems based on the autothermal reforming of these kinds of fuels in cooperation with industry. This paper reports about this work. (orig.)

Developments in fuel manufacturing

International Nuclear Information System (INIS)

Ion, S.E.; Harrop, G.; Maricalva Gonzalez, J.

1995-01-01

The status of the investment and R and D programmes in the UK and Spanish fuel fabrication facilities is outlined. Due to a number of circumstances, BNFL and ENUSA have been in the forefront of capital investment, with associated commitment to engineering and scientific research and development. Carrying through this investment has allowed the embodiment of proven state of the art technologies in the design of fuel fabrication plants, with particular emphasis on meeting the future challenge of health and safety, and product quality, at an acceptable cost. ENUSA and BNFL currently supply fuel, not only to their respective 'home' markets but also to France, Belgium, Sweden, and Germany. Both organisations employ an International Business outlook and partake in focused and speculative R and D projects for the design and manufacture of nuclear fuel. (orig./HP)

Development of wood fuel delivery logistics; Puupolttoaineiden hankintalogistiikan kehittaeminen

Energy Technology Data Exchange (ETDEWEB)

Laitinen, H

1997-12-31

The main aim of the project is to model the energy wood business and total logistics in a certain large region. First, wood utilisation locations inside this area are examined; the most important ones are the wood processing factories, and the heating- and power plants. After that, wood potentials in the forests of the area are evaluated in sub-areas suitable in size and sufficiently detailed for further evaluations. For that purpose, the most valuable source data are forest management plans, up to ten years forward, on which basis the wood fuel potentials can be evaluated following sustainable development. In Finland there are extensive and detailed data bases storing forest information and it is possible to collect necessary data for a data base applicable to our calculations. In logistical sense it is important to know, by which delivery chains the economically best and desired results are achieved. The software prototype based on data base is modelled and developed at VTT Energy, for facilitating these planning activities. The starting point of the planning system in delivery logistics is the implementation of an easy tool for versatile planning so that with this tool model different delivery chains can be flexiblyed, create usage scenarios, make alternative examinations, and calculate impacts of different factors on energy wood amounts yielded and delivery costs. With planning system in delivery logistics we calculate production costs and amounts delivered to different utilisation locations. The system offers tools for definitions of utilisation locations, calculation parameters, and delivery chains

Development of wood fuel delivery logistics; Puupolttoaineiden hankintalogistiikan kehittaeminen

Energy Technology Data Exchange (ETDEWEB)

Laitinen, H.

1996-12-31

The main aim of the project is to model the energy wood business and total logistics in a certain large region. First, wood utilisation locations inside this area are examined; the most important ones are the wood processing factories, and the heating- and power plants. After that, wood potentials in the forests of the area are evaluated in sub-areas suitable in size and sufficiently detailed for further evaluations. For that purpose, the most valuable source data are forest management plans, up to ten years forward, on which basis the wood fuel potentials can be evaluated following sustainable development. In Finland there are extensive and detailed data bases storing forest information and it is possible to collect necessary data for a data base applicable to our calculations. In logistical sense it is important to know, by which delivery chains the economically best and desired results are achieved. The software prototype based on data base is modelled and developed at VTT Energy, for facilitating these planning activities. The starting point of the planning system in delivery logistics is the implementation of an easy tool for versatile planning so that with this tool model different delivery chains can be flexiblyed, create usage scenarios, make alternative examinations, and calculate impacts of different factors on energy wood amounts yielded and delivery costs. With planning system in delivery logistics we calculate production costs and amounts delivered to different utilisation locations. The system offers tools for definitions of utilisation locations, calculation parameters, and delivery chains

New developments in dry spent fuel storage

International Nuclear Information System (INIS)

Bonnet, C.; Chevalier, Ph.

2001-01-01

As shown in various new examples, HABOG facility (Netherlands), CERNAVODA (Candu - Romania), KOZLODUY (WWER - Bulgaria), CHERNOBYL ( RMBK - Ukraine), MAYAK (Spent Fuel from submarine and Icebreakers - Russia), recent studies allow to confirm the flexibility and performances of the CASCAD system proposed by SGN, both in safety and operability, for the dry storage of main kinds of spent fuel. The main features are: A multiple containment barrier system: as required by international regulation, 2 independent barriers are provided (tight canister and storage pit); Passive cooling, while the Fuel Assemblies are stored in an inert atmosphere and under conditions of temperature preventing from degradation of rod cladding; Sub-criticality controlled by adequate arrangements in any conditions; Safe facility meeting ICPR 60 Requirements as well as all applicable regulations (including severe weather conditions and earthquake); Safe handling operations; Retrievability of the spent fuel either during storage period or at the end of planned storage period (100 years); Future Decommissioning of the facility facilitated through design optimisation; Construction and operating cost-effectiveness. (author)

Measurement and Analysis Plan for Investigation of Spent-Fuel Assay Using Lead Slowing-Down Spectroscopy

International Nuclear Information System (INIS)

Smith, Leon E.; Haas, Derek A.; Gavron, Victor A.; Imel, G.R.; Ressler, Jennifer J.; Bowyer, Sonya M.; Danon, Y.; Beller, D.

2009-01-01

Under funding from the Department of Energy Office of Nuclear Energy's Materials, Protection, Accounting, and Control for Transmutation (MPACT) program (formerly the Advanced Fuel Cycle Initiative Safeguards Campaign), Pacific Northwest National Laboratory (PNNL) and Los Alamos National Laboratory (LANL) are collaborating to study the viability of lead slowing-down spectroscopy (LSDS) for spent-fuel assay. Based on the results of previous simulation studies conducted by PNNL and LANL to estimate potential LSDS performance, a more comprehensive study of LSDS viability has been defined. That study includes benchmarking measurements, development and testing of key enabling instrumentation, and continued study of time-spectra analysis methods. This report satisfies the requirements for a PNNL/LANL deliverable that describes the objectives, plans and contributing organizations for a comprehensive three-year study of LSDS for spent-fuel assay. This deliverable was generated largely during the LSDS workshop held on August 25-26, 2009 at Rensselaer Polytechnic Institute (RPI). The workshop itself was a prominent milestone in the FY09 MPACT project and is also described within this report.

New long-term plan of nuclear development and perspectives of nuclear fuel cycle policy

International Nuclear Information System (INIS)

Uchiyama, Yohji

2005-01-01

Japan's nuclear fuel cycle policy, recently issued as an interim report of the Council to Formulate the New Long-Term Nuclear Program of the Atomic Energy Commission, is summarized and briefly explained together with the concluding remarks from the sub-committee for discussing technical and economical problems on the spent nuclear fuels with the present state of the Rokkasho reprocessing plant in mind. As for the nuclear fuel treatment, the panel considered four scenarios: (1) total reprocessing (the reprocessing for spent fuel after an appropriate period of storage); (2) partial reprocessing (spent fuel is reprocessed, with direct disposal of any spent fuel in excess of reprocessing capacity); (3) total direct disposal (direct disposal of all spent fuel); and (4) temporary storage (spent fuel is temporarily stored, and in about 2060 a choice will be made about whether to reprocess it or directly dispose of it). These four scenarios were studied from various perspectives, namely: (1) ensuring safety; (2) energy security; (3) environmental compatibility; (4) economic efficiency; (5) nuclear nonproliferation; (6) technical feasibility; (7) social acceptance; (8) securing choices; (9) issues concerning change in policy; and (10) overseas trends. Regarding economic efficiency, the council in particular conducted detailed studies and reassessment of nuclear fuel cycle costs. Scenario 1 (total reprocessing) is about 0.5-0.7 yen/kWh higher than scenario 3 (total direct disposal). However, looking at the situation from the perspectives of energy security, that is the stable supply and moderate use of resources, and environmental compatibility, scenario 1 (total reprocessing) can be evaluated as superior to the other scenarios. And more importantly, if the fast-breeder reactor cycle is commercialized, this superiority increases considerably. (S. Ohno)

Radioactive waste management plan for the PBMR (Pty) Ltd fuel plant

International Nuclear Information System (INIS)

Makgae, Mosidi E.

2009-01-01

The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice. The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by utilising processes that minimize, reduce, recover and recycle without exposing employees, the public and the environment to unacceptable impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams, for example gaseous, liquid or solid wastes.

Advanced high throughput MOX fuel fabrication technology and sustainable development

International Nuclear Information System (INIS)

Krellmann, Juergen

2005-01-01

The MELOX plant in the south of France together with the La Hague reprocessing plant, are part of the two industrial facilities in charge of closing the nuclear fuel cycle in France. Started up in 1995, MELOX has since accumulated a solid know-how in recycling plutonium recovered from spent uranium fuel into MOX: a fuel blend comprised of both uranium and plutonium oxides. Converting recovered Pu into a proliferation-resistant material that can readily be used to power a civil nuclear reactor, MOX fabrication offers a sustainable solution to safely take advantage of the plutonium's high energy content. Being the first large-capacity industrial facility dedicated to MOX fuel fabrication, MELOX distinguishes itself from the first generation MOX plants with high capacity (around 200 tHM versus around 40 tHM) and several unique operational features designed to improve productivity, reliability and flexibility while maintaining high safety standards. Providing an exemplary reference for high throughput MOX fabrication with 1,000 tHM produced since start-up, the unique process and technologies implemented at MELOX are currently inspiring other MOX plant construction projects (in Japan with the J-MOX plant, in the US and in Russia as part of the weapon-grade plutonium inventory reduction). Spurred by the growing international demand, MELOX has embarked upon an ambitious production development and diversification plan. Starting from an annual level of 100 tons of heavy metal (tHM), MELOX demonstrated production capacity is continuously increasing: MELOX is now aiming for a minimum of 140 tHM by the end of 2005, with the ultimate ambition of reaching the full capacity of the plant (around 200 tHM) in the near future. With regards to its activity, MELOX also remains deeply committed to sustainable development in a consolidated involvement within AREVA group. The French minister of Industry, on August 26th 2005, acknowledged the benefits of MOX fuel production at MELOX: 'In

High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

International Nuclear Information System (INIS)

Martin, M.M.

1993-01-01

The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

BISON Software V&V Plan

Energy Technology Data Exchange (ETDEWEB)

R. L. Williamson; J. D. Hales; D. M. Perez; S. R. Novascone; G. Pastore

2014-07-01

The primary vision for the BISON development team is to deliver a nuclear fuel performance simulation tool that is used to provide a researcher or fuel designer with best estimate calculations of the highly coupled and nonlinear phenomena that govern nuclear fuel behavior. Accurately simulating nuclear fuel behavior is a challenging computational undertaking and verification and validation (V&V) play an important role in realizing this vision. The purpose of this V&V plan is to express the BISON team’s definition of the terms verification and validation, document what we have done regarding V&V, and outline what we plan to do.

Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

International Nuclear Information System (INIS)

Kryter, R.C.

1975-01-01

Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

Industrial Fuel Gas Demonstration Plant Program: environmental permit compliance plan

Energy Technology Data Exchange (ETDEWEB)

Bodamer, Jr., James W.; Bocchino, Robert M.

1979-11-01

This Environmental Permit Compliance Plan is intended to assist the Memphis Light, Gas and Water Division in acquiring the necessary environmental permits for their proposed Industrial Fuel Gas Demonstration Plant in a time frame consistent with the construction schedule. Permits included are those required for installation and/or operation of gaseous, liquid and solid waste sources and disposal areas. Only those permits presently established by final regulations are described. The compliance plan describes procedures for obtaining each permit from identified federal, state and local agencies. The information needed for the permit application is presented, and the stepwise procedure to follow when filing the permit application is described. Information given in this plan was obtained by reviewing applicable laws and regulations and from telephone conversations with agency personnel on the federal, state and local levels. This Plan also presents a recommended schedule for beginning the work necessary to obtain the required environmental permits in order to begin dredging operations in October, 1980 and construction of the plant in September, 1981. Activity for several key permits should begin as soon as possible.

Calculation of parameters for inspection planning and evaluation: mixed-oxide fuel fabrication facilities

International Nuclear Information System (INIS)

Reardon, P.T.; Mullen, M.F.

1982-08-01

As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities for mixed-oxide fuel fabrication facilities. There were four distinct efforts involved in this task. These were as follows: show the effect on a material balance verification of using two variables measurement methods in some strata; perform additional calculations for the reference facility described in STR-89; modify the INSPECT computer programs to be used as an after-inspection analysis tool, as well as a preinspection planning tool; provide written comments and explantations of text and graphs of the first draft of STR-89, Safeguards Considerations for Mixed-Oxide Fuel Element Fabrication Facilities, by W. Bahm, T. Shea, and D. Tolchenkov, System Studies Section, IAEA

Quarterly Progress Report Fuels Development Operation: October - December 1959

Energy Technology Data Exchange (ETDEWEB)

Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

1960-01-15

The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

Quarterly Progress Report Fuels Development Operation: January - March 1958

Energy Technology Data Exchange (ETDEWEB)

Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

1958-04-15

The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

Quarterly Progress Report Fuels Development Operation: July - September 1957

Energy Technology Data Exchange (ETDEWEB)

Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Wallace, W. P. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

1957-10-15

The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

Test plan for Series 2 spent fuel cladding containment credit tests

International Nuclear Information System (INIS)

Wilson, C.N.

1984-10-01

This test plan describes a second series of tests to be conducted by Westinghouse Hanford Company (WHC) to evaluate the effectiveness of breached cladding as a barrier to radionuclide release in the NNWSI-proposed geologic repository. These tests will be conducted at the Hanford Engineering Development Laboratory (HEDL). A first series of tests, initiated at HEDL during FY 1983, demonstrated specimen preparation and feasibility of the testing concept. The second series tests will be similar to the Series 1 tests with the following exceptions: NNWSI reference groundwater obtained from well J-13 will be used as the leachant instead of deionized water; fuel from a second source will be used; and certain refinements will be made in specimen preparation, sampling, and analytical procedures. 12 references, 5 figures, 5 tables

Japanese programme on the development of high duty fuel and related power ramping tests

International Nuclear Information System (INIS)

Mishima, Y.

1983-01-01

Power ramping tests hitherto planned or carried out in Japan can be classified into two categories: 1) the tests programme by private organizations on fuel behaviour under various conditions of power ramping, in participating international programmes; and 2) a partially government sponsored programme, which was officially inaugurated in 1981 under the title of High-Duty Fuel Development Programme. The latter has been carried out by the Nuclear Power Engineering Test Centre, based on the schedule decided by the MITI Committee (chaired by the author), for a period of 10 years. These programmes will be described with emphasis on the latter (National Programme). (author)

Japanese programme on the development of high duty fuel and related power ramping tests

Energy Technology Data Exchange (ETDEWEB)

Mishima, Y

1983-06-01

Power ramping tests hitherto planned or carried out in Japan can be classified into two categories: 1) the tests programme by private organizations on fuel behaviour under various conditions of power ramping, in participating international programmes; and 2) a partially government sponsored programme, which was officially inaugurated in 1981 under the title of High-Duty Fuel Development Programme. The latter has been carried out by the Nuclear Power Engineering Test Centre, based on the schedule decided by the MITI Committee (chaired by the author), for a period of 10 years. These programmes will be described with emphasis on the latter (National Programme). (author)

Used fuel disposition research and development roadmap - FY10 status.

Energy Technology Data Exchange (ETDEWEB)

Nutt, W. M. (Nuclear Engineering Division)

2010-10-01

the U.S. and other nations in understanding disposal concepts in different geologic media, but gaps in knowledge still exist. A principal aspect of concern to the UFDC as it considers the broad issues of siting a repository in different geologic media are the marked differences in the regulatory bases for assessing suitability and safety of a repository between the U.S. and other nations. Because the probability based - risked informed nature of the current U.S. regulations for high-level radioactive waste and spent nuclear fuel is sufficiently different from other regulations, information gained in previous studies, while useful, likely need to be supplemented to enable more convincing communication with the public, better defense of the numerical models, and stronger safety cases. Thus, it was recognized when the UFDC was established that there were readily identified disposal-related R&D opportunities to address knowledge gaps. An effort to document these research opportunities was a key component of Fiscal Year (FY) 2010 engineered system, natural system, and system-level modeling activities for a range of disposal environments. A principal contribution to identifying these gaps was a workshop held to gather perspectives from experts both within and external to the UFDC regarding R&D opportunities. In the planning for FY2010 it was expected that these activities would culminate with a UFDC research and development roadmap that would identify the knowledge gaps, discuss the R&D needed to fill these gaps, and prioritize the proposed R&D over both the near- and long-term. A number of knowledge gaps and needed R&D were identified and are discussed in this report. However, these preliminary R&D topics have not been evaluated in detail nor have they been prioritized to support future planning efforts. This will be completed in FY11 and the final UFDC Research and Development Roadmap will be completed. This report discusses proposed R&D topics in three areas related to

Used fuel disposition research and development roadmap - FY10 status

International Nuclear Information System (INIS)

Nutt, W.M.

2010-01-01

made in the U.S. and other nations in understanding disposal concepts in different geologic media, but gaps in knowledge still exist. A principal aspect of concern to the UFDC as it considers the broad issues of siting a repository in different geologic media are the marked differences in the regulatory bases for assessing suitability and safety of a repository between the U.S. and other nations. Because the probability based - risked informed nature of the current U.S. regulations for high-level radioactive waste and spent nuclear fuel is sufficiently different from other regulations, information gained in previous studies, while useful, likely need to be supplemented to enable more convincing communication with the public, better defense of the numerical models, and stronger safety cases. Thus, it was recognized when the UFDC was established that there were readily identified disposal-related R and D opportunities to address knowledge gaps. An effort to document these research opportunities was a key component of Fiscal Year (FY) 2010 engineered system, natural system, and system-level modeling activities for a range of disposal environments. A principal contribution to identifying these gaps was a workshop held to gather perspectives from experts both within and external to the UFDC regarding R and D opportunities. In the planning for FY2010 it was expected that these activities would culminate with a UFDC research and development roadmap that would identify the knowledge gaps, discuss the R and D needed to fill these gaps, and prioritize the proposed R and D over both the near- and long-term. A number of knowledge gaps and needed R and D were identified and are discussed in this report. However, these preliminary R and D topics have not been evaluated in detail nor have they been prioritized to support future planning efforts. This will be completed in FY11 and the final UFDC Research and Development Roadmap will be completed. This report discusses proposed R and D

CANDU flexible and economical fuel technology in China

Energy Technology Data Exchange (ETDEWEB)

Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

2014-07-01

Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

Trends and Developments for Fast Neutron Reactors and Related Fuel Cycles

International Nuclear Information System (INIS)

Carré, Frank

2013-01-01

• FR13 – A unique and dedicated framework to share updates on national programs of Fast Reactor developments, projects of new builds and plans for the future: - Near term projects of sodium and lead-alloy Fast Reactors; - Gen-IV visions of sodium-cooled and alternative types of Fast Neutron Reactors (GFR, LFR…). • FR13 – A special emphasis put on Fast Reactor Safety, Sustainability of nuclear fuel cycle and Young Generation perspective. • FR13 – A catalyst for further collaborations and alliances: - To share visions of goals and advisable options for future Fast Reactors and Nuclear Fuel Cycle; - To share cost of R&D and large demonstrations (safety, security, recycling); - To progress towards harmonized international standards; - To integrate national projects into a consistent international roadmap

1986 fuel cell seminar: Program and abstracts

Energy Technology Data Exchange (ETDEWEB)

None

1986-10-01

Ninety nine brief papers are arranged under the following session headings: gas industry's 40 kw program, solid oxide fuel cell technology, phosphoric acid fuel cell technology, molten carbonate fuel cell technology, phosphoric acid fuel cell systems, power plants technology, fuel cell power plant designs, unconventional fuels, fuel cell application and economic assessments, and plans for commerical development. The papers are processed separately for the data base. (DLC)

Hanford 200 Areas Development Plan

Energy Technology Data Exchange (ETDEWEB)

Rinne, C.A.; Daly, K.S.

1993-08-01

The purpose of the Hanford 200 Areas Development Plan (Development Plan) is to guide the physical development of the 200 Areas (which refers to the 200 East Area, 200 West Area, and 200 Area Corridor, located between the 200 East and 200 West Areas) in accordance with US Department of Energy (DOE) Order 4320.lB (DOE 1991a) by performing the following: Establishing a land-use plan and setting land-use categories that meet the needs of existing and proposed activities. Coordinating existing, 5-year, and long-range development plans and guiding growth in accordance with those plans. Establishing development guidelines to encourage cost-effective development and minimize conflicts between adjacent activities. Identifying site development issues that need further analysis. Integrating program plans with development plans to ensure a logical progression of development. Coordinate DOE plans with other agencies [(i.e., Washington State Department of Ecology (Ecology) and US Environmental Protection Agency (EPA)]. Being a support document to the Hanford Site Development Plan (DOE-RL 1990a) (parent document) and providing technical site information relative to the 200 Areas.

Hanford 200 Areas Development Plan

International Nuclear Information System (INIS)

Rinne, C.A.; Daly, K.S.

1993-08-01

The purpose of the Hanford 200 Areas Development Plan (Development Plan) is to guide the physical development of the 200 Areas (which refers to the 200 East Area, 200 West Area, and 200 Area Corridor, located between the 200 East and 200 West Areas) in accordance with US Department of Energy (DOE) Order 4320.lB (DOE 1991a) by performing the following: Establishing a land-use plan and setting land-use categories that meet the needs of existing and proposed activities. Coordinating existing, 5-year, and long-range development plans and guiding growth in accordance with those plans. Establishing development guidelines to encourage cost-effective development and minimize conflicts between adjacent activities. Identifying site development issues that need further analysis. Integrating program plans with development plans to ensure a logical progression of development. Coordinate DOE plans with other agencies [(i.e., Washington State Department of Ecology (Ecology) and US Environmental Protection Agency (EPA)]. Being a support document to the Hanford Site Development Plan (DOE-RL 1990a) (parent document) and providing technical site information relative to the 200 Areas

Coal-Based Oxy-Fuel System Evaluation and Combustor Development; Oxy-Fuel Turbomachinery Development for Energy Intensive Industrial Applications

Energy Technology Data Exchange (ETDEWEB)

Hollis, Rebecca

2013-03-31

Clean Energy Systems, Inc. (CES) partnered with the U.S. Department of Energy’s National Energy Technology Laboratory in 2005 to study and develop a competing technology for use in future fossil-fueled power generation facilities that could operate with near zero emissions. CES’s background in oxy-fuel (O-F) rocket technology lead to the award of Cooperative Agreement DE-FC26-05NT42645, “Coal-Based Oxy-Fuel System Evaluation and Combustor Development,” where CES was to first evaluate the potential of these O-F power cycles, then develop the detailed design of a commercial-scale O-F combustor for use in these clean burning fossil-fueled plants. Throughout the studies, CES found that in order to operate at competitive cycle efficiencies a high-temperature intermediate pressure turbine was required. This led to an extension of the Agreement for, “Oxy-Fuel Turbomachinery Development for Energy Intensive Industrial Applications” where CES was to also develop an intermediate-pressure O-F turbine (OFT) that could be deployed in O-F industrial plants that capture and sequester >99% of produced CO2, at competitive cycle efficiencies using diverse fuels. The following report details CES’ activities from October 2005 through March 2013, to evaluate O-F power cycles, develop and validate detailed designs of O-F combustors (main and reheat), and to design, manufacture, and test a commercial-scale OFT, under the three-phase Cooperative Agreement.

Development of fuel performance and thermal hydraulic technology

International Nuclear Information System (INIS)

Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

2000-03-01

Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

Fuels planning: science synthesis and integration; social issues fact sheet 15: Landscape change and aesthetics

Science.gov (United States)

Christine Esposito

2006-01-01

Fuels management produces changes in the landscape that can impact scenic beauty. If people do not consider a forest to be scenic, they may think that the low scenic quality is a result of poor management or ecological health. This fact sheet looks at the relevency of the effects of natural and human-caused landscape changes, when planning fuels management.

Development of the computer model by using LINGO for Electric System Expansion Planning

Energy Technology Data Exchange (ETDEWEB)

Lee, Man Ki; Kim, Seung Su [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-05-01

The main purpose of this study is to develop the electric system expansion planning model based on linear programming method and to describe how to use the model. The algorithm of linear programming is provided by LINGO. The decision variables are electric generation capacity per period and electricity generated by fuel types per year. Integer program is implemented in the model in order to consider indivisibility of electric capacities. The model also allows the constraint of CO{sub 2} to be considered in the planning. (Author) 4 refs.,.

Hanford Site Development Plan

International Nuclear Information System (INIS)

Hathaway, H.B.; Daly, K.S.; Rinne, C.A.; Seiler, S.W.

1993-05-01

The Hanford Site Development Plan (HSDP) provides an overview of land use, infrastructure, and facility requirements to support US Department of Energy (DOE) programs at the Hanford Site. The HSDP's primary purpose is to inform senior managers and interested parties of development activities and issues that require a commitment of resources to support the Hanford Site. The HSDP provides an existing and future land use plan for the Hanford Site. The HSDP is updated annually in accordance with DOE Order 4320.1B, Site Development Planning, to reflect the mission and overall site development process. Further details about Hanford Site development are defined in individual area development plans

Dry refabrication technology development of spent nuclear fuel

International Nuclear Information System (INIS)

Park, Geun Il; Lee, J. W.; Song, K. C.

2012-04-01

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

Dry refabrication technology development of spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Geun Il; Lee, J. W.; Song, K. C.; and others

2012-04-15

Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.

Siemens fuel gasification technology - solutions and developments

Energy Technology Data Exchange (ETDEWEB)

Hannemann, F.; Schingnitz, M.; Schmid, C. [Siemens Fuel Gasification Technology GmbH, Freiberg (Germany)

2007-07-01

In 2006, Siemens Power Generation Group acquired the GSP Gasification technology, and renamed it SFGT. The presentation reviews the technology and provides an update on current projects. The future plans for the development of the technology based on extensive experience and comprehensive development work gathered over many years and proven in a number of gasification plants is covered. SFGT operates, at its Freiberg facility, a 5 MWth pilot plant which was built to test prototype designs and to determine process conditions for various feed streams. An overview is given of the results of tests completed on a wide range of carbonaceous materials including all types of solid fuels from lignite to anthracite, as well as brown coal, oil, sludge or biomass, and low-temperature coke or petcoke. The technical focus of the paper is on the unique design features such as the cooling screen and alternative refractory lining, as well as the dense flow feeding system that allows the preferable use of lignite applications.

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, J. S.; Hong, H. D.; Kim, Y. H.

2001-03-01

Since the amount of the spent fuel rapidly increases, the current R and D activities are focused on the technology development related with the storage and utilization of the spent fuel. In this research, to provide such a technology, the mechanical head-end process has been developed. In detail, the swing and shock-free crane and the RCGLUD(Remote Cask Grappling and Lid Unbolting Device) were developed for the safe transportation of the spent fuel assembly, the LLW drum and the transportation cask. Also, the disassembly devices required for the head-end process were developed. This process consists of an assembly downender, a rod extractor, a rod cutter, a fuel decladding device, a skeleton compactor, a force-rectifiable manipulator for the abnormal spent fuel disassembly, and the gantry type telescopic transporter, etc. To provide reliability and safety of these devices, the 3 dimensional graphic design system is developed. In this system, the mechanical devices are modelled and their operation is simulated in the virtual environment using the graphic simulation tools. So that the performance and the operational mal-function can be investigated prior to the fabrication of the devices. All the devices are tested and verified by using the fuel prototype at the mockup facility

Fuel development at CERCA. Status of development - September 1984

International Nuclear Information System (INIS)

Fanjas, Y.; Dewez, Ph.; Savornin, B.

1985-01-01

Since 1978, CERCA has developed high density aluminide (UAl x ), oxide (U 3 O 8 ) and silicides (U 3 Si 2 , U 3 Si) fuels allowing the use of 19.75 enriched uranium in research and test reactors. An extensive irradiation program has been carried out to test the full size fuel plates and fuel elements fabricated by CERCA. So far, all the irradiation tests have given satisfactory results whatever the uranium density, the burn-up level and the type of fuel. In particular, silicides which cover the whole density range from 1 to 7 g U/cm 3 appear more and more as the standard fuels for the future. (author)

Highlights of 50 years of nuclear fuels developments

International Nuclear Information System (INIS)

Simnad, M.T.

1989-01-01

The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel cycle costs and reliable performance in the fuel elements are discussed in the historical context

Proposal of a nuclear cycle research and development plan in Tokai works. The roadmap from LWR cycle to FBR cycle

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Abe, Tomoyuki; Kashimura, Takuo; Nagai, Toshihisa; Maeda, Seichiro; Yamaguchi, Toshiya; Kuroki, Ryoichiro

2003-07-01

The Generation-II Project Task Force Team has investigated a research and development plan of a future nuclear fuel cycle in Tokai works for about three months from December 19, 2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R and D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. Establishment and advancement of 'the tough nuclear fuel cycle'. Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste. And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology. (author)

Fuel development studies

International Nuclear Information System (INIS)

Michel, F.

1986-12-01

This paper describes the main lines of the studies carried out to develop the Fast Neutron Fuel Element, from the ''SPX1-first load'' version, to progress to high performance which will be required for the project 1500 and for the fast neutron series [fr

Strategies in development of advanced fuels for LMFBR

International Nuclear Information System (INIS)

Handa, Muneo

1976-12-01

Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Fuel alcohol opportunities for Indiana

Energy Technology Data Exchange (ETDEWEB)

Greenglass, Bert

1980-08-01

Prepared at the request of US Senator Birch Bayh, Chairman of the National Alcohol Fuels Commission, this study may be best utilized as a guidebook and resource manual to foster the development of a statewide fuel alcohol plan. It examines sectors in Indiana which will impact or be impacted upon by the fuel alcohol industry. The study describes fuel alcohol technologies that could be pertinent to Indiana and also looks closely at how such a fuel alcohol industry may affect the economic and policy development of the State. Finally, the study presents options for Indiana, taking into account the national context of the developing fuel alcohol industry which, unlike many others, will be highly decentralized and more under the control of the lifeblood of our society - the agricultural community.

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

Energy Technology Data Exchange (ETDEWEB)

Matloch, L.; Vaccaro, S.; Couland, M.; De Baere, P.; Schwalbach, P. [Euratom, Communaute europeenne de l' energie atomique - CEEA (European Commission (EC))

2015-07-01

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

Technology developments for Japanese BWR MOX fuel utilization

International Nuclear Information System (INIS)

Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

1997-01-01

The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

Standard format and content for radiological contingency plans for fuel cycle and materials facilities. Regulatory report

International Nuclear Information System (INIS)

1981-07-01

This report is issued as guidance to those fuel cycle and major materials licensees who are required by the NRC to prepare and submit a radiological contingency plan. This Standard Format has been prepared to help assure uniformity and completeness in the preparation of those plans

Development of MOX fuel database

International Nuclear Information System (INIS)

Ikusawa, Yoshihisa; Ozawa, Takayuki

2007-03-01

We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

Definition of Technology Readiness Levels for Transmutation Fuel Development

International Nuclear Information System (INIS)

Jon Carmack; Kemal O. Pasamehmetoglu

2008-01-01

To quantitatively assess the maturity of a given technology, the Technology Readiness Level (TRL) process is used. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Transmutation fuel development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the transmutation fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Transuranic Fuel Development Campaign

Spent fuel storage process equipment development

International Nuclear Information System (INIS)

Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

1990-02-01

Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)

Development of nuclear fuel materials for research reactor

International Nuclear Information System (INIS)

Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

1997-09-01

This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

Highlights of 50 years of nuclear fuel development

International Nuclear Information System (INIS)

Simnad, M.T.

1989-01-01

The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel-cycle costs and reliable performance in the fuel elements are discussed in the historical context. 10 refs

Developing Custom Fire Behavior Fuel Models for Mediterranean Wildland-Urban Interfaces in Southern Italy

Science.gov (United States)

Elia, Mario; Lafortezza, Raffaele; Lovreglio, Raffaella; Sanesi, Giovanni

2015-09-01

The dramatic increase of fire hazard in wildland-urban interfaces (WUIs) has required more detailed fuel management programs to preserve ecosystem functions and human settlements. Designing effective fuel treatment strategies allows to achieve goals such as resilient landscapes, fire-adapted communities, and ecosystem response. Therefore, obtaining background information on forest fuel parameters and fuel accumulation patterns has become an important first step in planning fuel management interventions. Site-specific fuel inventory data enhance the accuracy of fuel management planning and help forest managers in fuel management decision-making. We have customized four fuel models for WUIs in southern Italy, starting from forest classes of land-cover use and adopting a hierarchical clustering approach. Furthermore, we provide a prediction of the potential fire behavior of our customized fuel models using FlamMap 5 under different weather conditions. The results suggest that fuel model IIIP (Mediterranean maquis) has the most severe fire potential for the 95th percentile weather conditions and the least severe potential fire behavior for the 85th percentile weather conditions. This study shows that it is possible to create customized fuel models directly from fuel inventory data. This achievement has broad implications for land managers, particularly forest managers of the Mediterranean landscape, an ecosystem that is susceptible not only to wildfires but also to the increasing human population and man-made infrastructures.

Developing Custom Fire Behavior Fuel Models for Mediterranean Wildland-Urban Interfaces in Southern Italy.

Science.gov (United States)

Elia, Mario; Lafortezza, Raffaele; Lovreglio, Raffaella; Sanesi, Giovanni

2015-09-01

The dramatic increase of fire hazard in wildland-urban interfaces (WUIs) has required more detailed fuel management programs to preserve ecosystem functions and human settlements. Designing effective fuel treatment strategies allows to achieve goals such as resilient landscapes, fire-adapted communities, and ecosystem response. Therefore, obtaining background information on forest fuel parameters and fuel accumulation patterns has become an important first step in planning fuel management interventions. Site-specific fuel inventory data enhance the accuracy of fuel management planning and help forest managers in fuel management decision-making. We have customized four fuel models for WUIs in southern Italy, starting from forest classes of land-cover use and adopting a hierarchical clustering approach. Furthermore, we provide a prediction of the potential fire behavior of our customized fuel models using FlamMap 5 under different weather conditions. The results suggest that fuel model IIIP (Mediterranean maquis) has the most severe fire potential for the 95th percentile weather conditions and the least severe potential fire behavior for the 85th percentile weather conditions. This study shows that it is possible to create customized fuel models directly from fuel inventory data. This achievement has broad implications for land managers, particularly forest managers of the Mediterranean landscape, an ecosystem that is susceptible not only to wildfires but also to the increasing human population and man-made infrastructures.

A Collaborative Approach to Transportation Planning: Federal and State Perspectives on Section 180(c) Program Development

International Nuclear Information System (INIS)

Macaluso, C.; Strong, T.; Janairo, L.; Helvey, E.

2006-01-01

The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) committed in its November 2003 Strategic Plan for the Safe Transportation of Spent Nuclear Fuel and High-Level Radioactive Waste to Yucca Mountain: A Guide to Stakeholder Interactions to develop the transportation system collaboratively with stakeholders. The Strategic Plan further stated that four state regional groups (SRGs) would be the 'anchors' for OCRWM's collaboration with the states. The first major transportation planning activity that OCRWM initiated after publication of the Strategic Plan was the development of the Section 180(c) grant program. This document describes that collaboration and its outcomes from the perspective of the OCRWM participants and one of the SRGs, the Council of State Governments - Midwestern Office (CSG Midwest). (authors)

Development of TVSA VVER-1000 fuel

International Nuclear Information System (INIS)

Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

2013-01-01

The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

Irradiation Planning for Fully-Ceramic Micro-encsapsulated fuel in ATR at LWR-relevant conditions: year-end report on FY-2011

International Nuclear Information System (INIS)

Ougouag, Abderrafi M.; Sen, R. Sonat; Pope, Michael A.; Boer, Brian

2011-01-01

This report presents the estimation of required ATR irradiation levels for the DB-FCM fuel design (fueled with Pu and MAs). The fuel and assembly designs are those considered in a companion report [R. S. Sen et al., FCRandD-2011- 00037 or INL/EXT-11-23269]. These results, pertaining to the DB-FCM fuel, are definitive in as much as the design of said fuel is definitive. In addition to the work performed, as required, for DB-FCM fuel, work has started in a preliminary fashion on single-cell UO2 and UN fuels. These latter activities go beyond the original charter of this project and although the corresponding work is incomplete, significant progress has been achieved. However, in this context, all that has been achieved is only preliminary because the corresponding fuel designs are neither finalized nor optimized. In particular, the UO2 case is unlikely to result in a viable fuel design if limited to enrichment at or under 20 weight % in U-235. The UN fuel allows reasonable length cycles and is likely to make an optimal design possible. Despite being limited to preliminary designs and offering only preliminary conclusions, the irradiation planning tasks for UO2 and UN fuels that are summarized in this report are useful to the overall goal of devising and deploying FCM-LWR fuel since the methods acquired and tested in this project and the overall procedure for planning will be available for planning tests for the finalized fuel design. Indeed, once the fuel design is finalized and the expected burnup level is determined, the methodology that has been assembled will allow the prompt finalization of the neutronic planning of the irradiation experiment and would provide guidance on the expected experimental performance of the fuel. Deviations from the expected behavior will then have to be analyzed and the outcome of the analysis may be corrections or modifications for the assessment models as well as, possibly, fuel design modifications, and perhaps even variation of

Reprocessing in breeder fuel cycles

International Nuclear Information System (INIS)

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

1982-01-01

Over the past decade, the United States has developed plans and carried out programs directed toward the demonstration of breeder fuel reprocessing in connection with the first breeder demonstration reactor. A renewed commitment to moving forward with the construction of the Clinch River Breeder Reactor (CRBR) has been made, with startup anticipated near the end of this decade. While plans for the CRBR and its associated fuel cycle are still being firmed up, the basic research and development programs required to carry out the demonstrations have continued. This paper updates the status of the reprocessing plans and programs. Policies call for breeder recycle to begin in the early to mid-1990's. Contents of this paper are: (1) evolving plans for breeder reprocessing (demonstration reprocessing plant, reprocessing head-end colocated at an existing facility); (2) relationship to LWR reprocessing; (3) integrated equipment test (IET) facility and related hardware development activities (mechanical considerations in shearing and dissolving, remote operations and maintenance demonstration phase of IET, integrated process demonstration phase of IET, separate component development activities); and (4) supporting process R and D

The Canadian CANDU fuel development program and recent fuel operating experience

International Nuclear Information System (INIS)

Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

1999-01-01

This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the two-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These two programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

The conditions of gaseous fuels development

International Nuclear Information System (INIS)

Anon.

1996-01-01

Face to the actual situation of petrol and gas oil in France, the situation of gaseous fuels appears to be rather modest. However, the aim of gaseous fuels is not to totally supersede the liquid fuels. Such a situation would imply a complete overturn which has not been seriously considered yet. This short paper describes the essential conditions to promote the wider use of gaseous fuels: the intervention of public authorities to adopt a more advantageous tax policy in agreement with the ''Clean Air''law project, a suitable distribution network for gaseous fuels, a choice of vehicles consistent with the urban demand, the development of transformation kits of quality and of dual-fuel vehicles by the car manufacturers. (J.S.)

Hanford Site Development Plan

International Nuclear Information System (INIS)

Hathaway, H.B.; Daly, K.S.; Rinne, C.A.; Seiler, S.W.

1992-05-01

The Hanford Site Development Plan (HSDP) provides an overview of land use, infrastructure, and facility requirements to support US Department of Energy (DOE) programs at the Hanford Site. The HSDP's primary purpose is to inform senior managers and interested parties of development activities and issues that require a commitment of resources to support the Hanford Site. The HSDP provides a land use plan for the Hanford Site and presents a picture of what is currently known and anticipated in accordance with DOE Order 4320.1B. Site Development Planning. The HSDP wig be updated annually as future decisions further shape the mission and overall site development process. Further details about Hanford Site development are defined in individual area development plans

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.)

Development of spent fuel remote handling technology

International Nuclear Information System (INIS)

Yoon, J. S.; Hong, H. D.; Kim, S. H.

2004-02-01

In this research, the remote handling technology is developed for the advanced spent fuel conditioning process which gives a possible solution to deal with the rapidly increasing spent fuels. In detail, a fuel rod slitting device is developed for the decladding of the spent fuel. A series of experiments has been performed to find out the optimal condition of the spent fuel voloxidation which converts the UO 2 pellet into U 3 O 8 powder. The design requirements of the ACP equipment for hot test is established by analysing the modular requirement, radiation hardening and thermal protection of the process equipment, etc. The prototype of the servo manipulator is developed. The manipulator has an excellent performance in terms of the payload to weight ratio that is 30 % higher than that of existing manipulators. To provide reliability and safety of the ACP, the 3 dimensional graphic simulator is developed. Using the simulator the remote handling operation is simulated and as a result, the optimal layout of ACP is obtained. The supervisory control system is designed to control and monitor the several different unit processes. Also the failure monitoring system is developed to detect the possible accidents of the reduction reactor

Greater use of wood residue fuels through improved financial planning: a case study in Alabama

Energy Technology Data Exchange (ETDEWEB)

Billings, C.D.; Ziemke, M.C. (Alabama Univ., Huntsville, AL (United States). Coll. of Administrative Science); Stanford, R. (Alabama Dept. of Economic and Community Affairs, Montgomery, AL (United States))

1993-01-01

As the world reacts to environmental concerns relating to fossil energy usage, emphasis is again placed on greater use of renewable fuels such as wood residues. Realistically, however, decisions to utilize such fuels are based on economic factors, rather than desires to improve US energy independence and/or protect the environment. Because Alabama has a large forest products industry, state authorities have long sought to assist potential users of wood residue fuels to better use biomass fuels instead of the usual alternative: natural gas. State agency experience in promoting commercial and industrial use of wood residue fuels has shown that inadequate financial planning has often resulted in rejection of viable projects or acceptance of non-optimum projects. This paper discusses the reasons for this situation and suggests remedies for its improvement. (author)

Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

1992-05-01

This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

Workshop on IAEA Tools for Nuclear Energy System Assessment for Long-Term Planning and Development

International Nuclear Information System (INIS)

2009-01-01

The purpose of the workshop is to present to Member States tools and methods that are available from the IAEA in support of long-term energy planning and nuclear energy system assessments, both focusing on the sustainable development of nuclear energy. This includes tools devoted to energy system planning, indicators for sustainable energy development, the INPRO methodology for Nuclear Energy System Assessment (NESA) and tools for analysing nuclear fuel cycle material balance. The workshop also intends to obtain feedback from Member States on applying the tools, share experiences and lessons learned, and identify needs for IAEA support

Development of oxygen scavenger additives for jet fuels

Energy Technology Data Exchange (ETDEWEB)

Beaver, B.D.; Demunshi, R.; Sharief, V.; Tian, D.; Teng, Y. [Duquesne Univ., Pittsburgh, PA (United States)

1995-05-01

Our current research program is in response to the US Air Force`s FY93 New Initiative entitled {open_quotes}Advanced Fuel Composition and Use.{close_quotes} The critical goal of this initiative is to develop aircraft fuels which can operate at supercritical conditions. This is a vital objective since future aircraft designs will transfer much higher heat loads into the fuel as compared with current heat loads. In this paper it is argued that the thermal stability of most jet fuels would be dramatically improved by the efficient in flight-removal of a fuel`s dissolved oxygen. It is proposed herein to stabilize the bulk fuel by the addition of an additive which will be judiciously designed and programmed to react with oxygen and produce an innocuous product. It is envisioned that a thermally activated reaction will occur, between the oxygen scavenging additive and dissolved oxygen, in a controlled and directed manner. Consequently formation of insoluble thermal degradation products will be limited. It is believed that successful completion of this project will result in the development of a new type of jet fuel additive which will enable current conventional jet fuels to obtain sufficient thermal stability to function in significantly higher temperature regimes. In addition, it is postulated that the successful development of thermally activated oxygen scavengers will also provide the sub-critical thermal stability necessary for future development of endothermic fuels.

Status of research reactor fuel development in KAERI

International Nuclear Information System (INIS)

Kim, Chang-Kyu; Ryu, Woo-Seok; Park, Jong-Man; Lee, Don-Bae; Kim, Ki-Hwan; Kuk, Il-Hyun

1996-01-01

The development of uranium silicide dispersion fuel fabrication technology has been carried out in KAERI. LEU fuel bundle was prepared for irradiation test. In order to compare the performance of atomized and comminuted U 3 Si dispersed fuels, the bundle of two kinds of fuel elements were prepared. Irradiation test will be performed in the OR-hole of HANARO in the near future. U 3 Si 2 atomization technology has been improved by using ceramic crucible and nozzle. Irradiation test for atomized U 3 Si 2 plate type fuel will be carried out in cooperation with ANL by using HANARO in connection with RERTR advanced fuel development. (author)

Development on nuclear fuel cycle business in Japan

International Nuclear Information System (INIS)

Usami, Kogo

2002-01-01

The Japan Nuclear Fuel Co., Ltd. (JNF) develops five businesses on nuclear fuel cycle such as uranium concentration, storage and administration of high level radioactive wastes, disposition of low level radioactive wastes, used fuel reprocessing, MOX fuel, at Rokkasho-mura in Aomori prefecture. Here were introduced on outline, construction and operation in reprocessing and MOX fuel works, outline, present state and future subjects on technical development of uranium concentration, outline and safety of disposition center on low level radioactive wastes, and storage and administration of high level radioactive wastes. (G.K.)

1994 Site Development Plan: A plan with vision

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

The 1994 Lawrence Livermore National Laboratory Site Development Plan has been developed during a period of great change and uncertainty. Our goal is to make possible the best use of the Laboratory`s resources to meet shifting national priorities in the post-Cold War world. Site Planning is an important component of the overall Laboratory strategic planning process. This plan focuses on opportunities for the Laboratory as well as on key site development issues including facility construction, redevelopment and reuse, site accessibility, and security. A major challenge is to achieve sufficient stability in the site planning and execution so that the processes of construction can occur efficiently while at the same time providing sufficient flexibility in site facilities so that a range of changing national needs can be accommodated. We are closely coupled to the DOE strategic planning process to meet this challenge.

BE (fuel element)/ZL (interim storage facility) module. Constituents of the fuel BE data base for BE documentation with respect to the disposal planning and the support of the BE container storage administration

International Nuclear Information System (INIS)

Hoffmann, V.; Deutsch, S.; Busch, V.; Braun, A.

2012-01-01

The securing of spent fuel element disposal from German nuclear power plants is the main task of GNS. This includes the container supply and the disposal analysis and planning. Therefore GNS operates a data base comprising all in Germany implemented fuel elements and all fuel element containers in interim storage facilities. With specific program modules the data base serves an optimized repository planning for all spent fuel elements from German NPPS and the supply of required data for future final disposal. The data base has two functional models: the BE (fuel element) and the ZL (interim storage) module. The contribution presents the data structure of the modules and details of the data base operation.

Feasibility study on tandem fuel cycle

International Nuclear Information System (INIS)

Han, P.S.; Suh, I.S.; Rim, C.S.; Kim, B.K.; Suh, K.S.; Ro, S.K.; Juhn, P.I.; Kim, S.Y.

1983-01-01

The objective of this feasibility study is to review and assess the current state of technology concerning the tandem fuel cycle. Based on the results from this study, a long-term development plan suitable for Korea has been proposed for this cycle, i.e., the PWR → CANDU tandem fuel cycle which used plutonium and uranium, recovered from spent PWR fuel by co-processing, as fuel material for CANDU reactors. (Author)

Hydrogen-enriched fuels

Energy Technology Data Exchange (ETDEWEB)

Roser, R. [NRG Technologies, Inc., Reno, NV (United States)

1998-08-01

NRG Technologies, Inc. is attempting to develop hardware and infrastructure that will allow mixtures of hydrogen and conventional fuels to become viable alternatives to conventional fuels alone. This commercialization can be successful if the authors are able to achieve exhaust emission levels of less than 0.03 g/kw-hr NOx and CO; and 0.15 g/kw-hr NMHC at full engine power without the use of exhaust catalysts. The major barriers to achieving these goals are that the lean burn regimes required to meet exhaust emissions goals reduce engine output substantially and tend to exhibit higher-than-normal total hydrocarbon emissions. Also, hydrogen addition to conventional fuels increases fuel cost, and reduces both vehicle range and engine output power. Maintaining low emissions during transient driving cycles has not been demonstrated. A three year test plan has been developed to perform the investigations into the issues described above. During this initial year of funding research has progressed in the following areas: (a) a cost effective single-cylinder research platform was constructed; (b) exhaust gas speciation was performed to characterize the nature of hydrocarbon emissions from hydrogen-enriched natural gas fuels; (c) three H{sub 2}/CH{sub 4} fuel compositions were analyzed using spark timing and equivalence ratio sweeping procedures and finally; (d) a full size pick-up truck platform was converted to run on HCNG fuels. The testing performed in year one of the three year plan represents a baseline from which to assess options for overcoming the stated barriers to success.

Status and development of the thorium fuel cycle

International Nuclear Information System (INIS)

Yi Weijing; Wei Renjie

2003-01-01

A perspective view of the thorium fuel cycle is provided in this paper. The advantages and disadvantages of the thorium fuel cycle are given and the development of thorium fuel cycle in several types of reactors is introduced. The main difficulties in developing the thorium fuel cycle lie in the reprocessing and disposal of the waste and its economy, and the ways tried by foreign countries to solve the problems are presented in the paper

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

Endplug Welding Techniques developed for SFR Metallic Fuel Elements

International Nuclear Information System (INIS)

Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

2013-01-01

In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

An alternative format for Category I fuel cycle facility physical protection plans

International Nuclear Information System (INIS)

Dwyer, P.A.

1992-06-01

This document provides an alternative format for physical protection plans designed to meet the requirements of Title 10 of the Code of Federal Regulations, Sections 73.20, 73.45, and 73.46. These requirements apply to licensees who operate Category I fuel cycle facilities. Such licensees are authorized to use or possess a formula quantity of strategic special nuclear material. The format described is an alternative to that found under Regulatory Guide 5.52, Rev. 2 ''Standard Format and Content of a Licensee Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites (Other than Nuclear Power Plants).''

Research and development of thorium fuel cycle

International Nuclear Information System (INIS)

Oishi, Jun.

1994-01-01

Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

Advanced PEFC development for fuel cell powered vehicles

Science.gov (United States)

Kawatsu, Shigeyuki

Vehicles equipped with fuel cells have been developed with much progress. Outcomes of such development efforts include a Toyota fuel cell electric vehicle (FCEV) using hydrogen as the fuel which was developed and introduced in 1996, followed by another Toyota FCEV using methanol as the fuel, developed and introduced in 1997. In those Toyota FCEVs, a fuel cell system is installed under the floor of each RAV4L, to sports utility vehicle. It has been found that the CO concentration in the reformed gas of methanol reformer can be reduced to 100 ppm in wide ranges of catalyst temperature and gas flow rate, by using the ruthenium (Ru) catalyst as the CO selective oxidizer, instead of the platinum (Pt) catalyst known from some time ago. It has been also found that a fuel cell performance equivalent to that with pure hydrogen can be ensured even in the reformed gas with the carbon monoxide (CO) concentration of 100 ppm, by using the Pt-Ru (platinum ruthenium alloy) electrocatalyst as the anode electrocatalyst of a polymer electrolyte fuel cell (PEFC), instead of the Pt electrocatalyst known from some time ago.

Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

Energy Technology Data Exchange (ETDEWEB)

Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

1994-07-01

The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.